ML25038A131
| ML25038A131 | |
| Person / Time | |
|---|---|
| Issue date: | 01/15/2025 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | |
| References | |
| NRC-0183 | |
| Download: ML25038A131 (1) | |
Text
Official Transcript of Proceedings NUCLEAR REGULATORY COMMISSION
Title:
Advisory Committee on Reactor Safeguards NuScale Design-Centered Review Open Session Docket Number:
(n/a)
Location:
teleconference Date:
Wednesday, January 15, 2025 Work Order No.:
NRC-0183 Pages 1-93 NEAL R. GROSS AND CO., INC.
Court Reporters and Transcribers 1716 14th Street, N.W.
Washington, D.C. 20009 (202) 234-4433
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 1
1 2
3 DISCLAIMER 4
5 6
UNITED STATES NUCLEAR REGULATORY COMMISSIONS 7
ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 8
9 10 The contents of this transcript of the 11 proceeding of the United States Nuclear Regulatory 12 Commission Advisory Committee on Reactor Safeguards, 13 as reported herein, is a record of the discussions 14 recorded at the meeting.
15 16 This transcript has not been reviewed, 17 corrected, and edited, and it may contain 18 inaccuracies.
19 20 21 22 23
1 UNITED STATES OF AMERICA 1
NUCLEAR REGULATORY COMMISSION 2
+ + + + +
3 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 4
(ACRS) 5
+ + + + +
6 NUSCALE DESIGN-CENTERED SUBCOMMITTEE 7
+ + + + +
8 WEDNESDAY 9
JANUARY 15, 2025 10
+ + + + +
11 The Subcommittee met via Teleconference, 12 at 8:30 a.m. EST, Walter L. Kirchner, Chair, 13 presiding.
14 15 COMMITTEE MEMBERS:
16 WALTER L. KIRCHNER, Chair 17 RONALD G. BALLINGER, Member 18 VICKI M. BIER, Member 19 VESNA B. DIMITRIJEVIC, Member 20 CRAIG A. HARRINGTON, Member 21 GREGORY H. HALNON, Member 22 ROBERT P. MARTIN, Member 23 SCOTT P. PALMTAG, Member 24 DAVID A. PETTI, Member 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
2 THOMAS E. ROBERTS, Member 1
MATTHEW W. SUNSERI, Member 2
3 ACRS CONSULTANTS:
4 DENNIS BLEY 5
7 DESIGNATED FEDERAL OFFICIAL:
10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
3 CONTENTS 1
Opening Remarks.................
4 2
Discussion of SDAA Chapter 3 and 16....... 12 3
Staff's Evaluation of NuScale SDAA 4
Chapter 16 (NRR)
................ 19 5
Discussion of NuScale LOCA Evaluation Model 6
Topical Report and Status of High-Impact 7
Technical Issues
................ 61 8
Staff's Evaluation of NuScale LOCA Evaluation 9
Model Topical Report
.............. 79 10 Public Comment
................. 92 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
4 P-R-O-C-E-E-D-I-N-G-S 1
8:30 a.m.
2 CHAIR KIRCHNER: The meeting will now come 3
to order. This is a meeting of the NuScale Design-4 Centered Review Subcommittee of the Advisory Committee 5
on Reactor Safeguards.
6 I am Walt Kirchner, chair of today's 7
subcommittee meeting. ACRS members in attendance in 8
person are Ron Ballinger, Greg
- Halnon, Craig 9
Harrington, Bob Martin, Scott Palmtag, and Tom 10 Roberts. ACRS members in attendance virtually via 11 Teams are Vicki Bier, Vesna Dimitrijevic, David Petti, 12 Matt Sunseri, and myself. We have one of our 13 consultants participating in-person, Steve Schultz, 14 and one of our consultants participating virtually via 15 Teams, Dennis Bley. If I have missed anyone, either 16 ACRS members or consultants, please speak up now.
17 (No response.)
18 CHAIR KIRCHNER: Michael Snodderly of the 19 ACRS staff is the Designated Federal Officer for this 20 meeting.
21 No member conflicts of interest were 22 identified for today's meeting, and I note that we 23 have a quorum, as well.
24 During today's meeting, this Subcommittee 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
5 will receive a briefing on the staff's evaluation of 1
NuScale Power LLC's US460 Standard Design Approval 2
Application; Chapter 3, Design of Structures, Systems, 3
Components, and Equipment; Chapter 16, Technical 4
Specifications; and Loss-of-Coolant Accident 5
Evaluation Model Topical Report. We will also be 6
briefed on the status of high-impact technical issues 7
by the NuScale staff.
8 We previously reviewed the certified 9
NuScale US600 design, as documented in our July 29, 10 2020 letter report, "Report on the Safety Aspects of 11 the NuScale Small Modular Reactor." Like the staff, 12 we are performing a delta review between the two 13 designs, including a power uprate from 50 to 77 14 megawatts electric per module.
15 We are reviewing these chapters as part of 16 our statutory obligation under Title 10 of the Code of 17 Federal Regulations, Part 52, Subpart E, Section 141, 18 referral to the Advisory Committee on Reactor 19 Safeguards to report on those portions of the 20 application which concern safety.
21 The ACRS was established by statute and is 22 governed by the Federal Advisory Committee Act, or 23 FACA. The NRC implements FACA in accordance with our 24 regulations.
Per these regulations and the 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
6 Committee's bylaws, the ACRS speaks only through its 1
published letter reports. All member comments should 2
be regarded as only the individual opinion of that 3
member and not a Committee position.
4 All relevant information related to ACRS 5
activities, such as letters, rules for meeting 6
participation, and transcripts are located in the NRC 7
public website and can be easily found by typing 8
"about us ACRS" in the search field on NRC's homepage.
9 The ACRS, consistent with the agency's 10 value of public transparency and regulation of nuclear 11 facilities, provides opportunity for public input and 12 comment during our proceedings. We have received no 13 written statements or a request to make an oral 14 statement from the public, but we have set aside time 15 at the end of this meeting for public comments.
16 Portions of this meeting may be closed to 17 protect sensitive information, as required by FACA and 18 the Government in the Sunshine Act. Attendance during 19 the closed portion of the meeting will be limited to 20 the NRC staff and its consultants, applicants, and 21 those individuals and/or organizations who have 22 entered into an appropriate confidentiality agreement.
23 We will confirm that only eligible individuals are in 24 the closed portion of the meeting.
25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
7 The ACRS will gather information, analyze 1
the relevant issues and facts, and formulate proposed 2
conclusions and recommendations, as appropriate, for 3
deliberation by the full Committee.
4 A transcript of the meeting is being kept 5
and will be posted on our website. When addressing 6
the Subcommittee, the participants should first 7
identify themselves, and speak with sufficient clarity 8
and volume so that they may be readily heard. If you 9
are not speaking, please mute your computer on Teams, 10 or by pressing star-6 if you're on your phone.
11 Please do not use the Teams chat feature 12 to conduct sidebar discussions related to the 13 presentations; rather, limit the use of that function 14 to report IT problems. We ask everyone in the room, 15 please put all your electronic devices on silent mode 16 and mute your laptop microphone and speakers. In 17 addition, please keep sidebar discussions in the room 18 to a minimum, since the ceiling microphones are live.
19 For the presenters, your table microphones 20 are unidirectional and you'll need to speak into the 21 front of the microphone to be heard.
22 Finally, if you have any feedback for the 23 ACRS about today's meeting, we encourage you to fill 24 out the public meeting feedback form on the NRC's 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
8 website.
1 And now I just want to make a personal 2
note. Here in Santa Fe, it's a nice, balmy 15 3
degrees. It's dark. We have a full moon out. On the 4
West Coast, where our colleagues from NuScale are, 5
it's probably much darker out there. I hope the 6
moon's out, it's spectacular.
7 Thank you. We normally try and schedule 8
the NuScale meetings for afternoon sessions, but, 9
given the amount of material we want to cover today, 10 NuScale has agreed to join us at -- I think it's about 11 5:30 a.m. out there in Oregon.
12 So, with that, thank you. And we'll now 13 proceed with the meeting. I think we'll start with a 14 opening statement from the NRC staff.
15 Greg, if you could direct things from 16 there?
17 MEMBER HALNON: Go ahead.
18 MR.
JARDANE:
Good
- morning, Chair 19 Kirchner, Vice Chair Halnon. Good morning to the ACRS 20 Subcommittee members, NuScale participants, NRC staff, 21 and members of the public. My name is Mahmoud 22 Jardane, and I serve as the branch chief of the New 23 Reactor Licensing
- Branch, responsible for the 24 licensing of NuScale US460 design, in the Division of 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
9 New and Renewed Licenses, NRR.
1 Thank you for the opportunity --
2 CHAIR KIRCHNER: Could I interrupt you for 3
a minute? Sorry. Would you pull your microphone 4
closer? Those of us on Teams are not getting a good 5
audio signal.
6 MR. JARDANE: All right, is that any 7
better?
8 (No response.)
9 MR.
JARDANE:
Thank you for the 10 opportunity today for the staff to present on their 11 review of select NuScale US Standard Design Approval 12 Application, SDAA, Chapters and topical reports. As 13 you are aware, the staff is reviewing all chapters of 14 the SDAA concurrently, with staggered completion dates 15 based on the complexity of the chapter and the extent 16 of the change from the certified NuScale US600 design.
17 Today, the staff will be presenting on 18 their review of the fourth group of SDAA chapters and 19 topical reports, including Chapter 16, Technical 20 Specifications, and the Loss-of-Coolant Accident 21 Evaluation Model Topical Report.
22 Earlier this year, the staff presented to 23 the Subcommittee on Chapter 2, portions of Chapter 3, 24 Chapters 7, 8, 9, 10, 11, 12, 13, and 14, portions of 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
10 Chapter 17, and Chapter 18. Staff is finalizing their 1
review of the remaining SDAA chapters and topical 2
reports and will inform the ACRS on the safety 3
evaluation of the remaining chapters. Topical reports 4
are available to the ACRS.
5 In today's meeting, the staff will focus 6
on the deltas from the design specification that the 7
NRC has approved and this Committee -- and the 8
Committee reviewed in the past. Once again, thank you 9
for the opportunity, and we look forward to a good 10 discussion.
11 MEMBER HALNON: Thank you, MJ.
12 CHAIR KIRCHNER: Thank you. Greg, if I 13 may, I noticed I made a mistake. We are not covering 14 Chapter 3 today. That will be taken up next month, on 15 February -- let me just get the date -- it will be on 16 Tuesday, February 4. So, just Chapter 16, the LOCA 17 Topical Report, and the HITI status. Thank you.
18 MR. TESFAYE: Excuse me. This is Getachew 19 Tesfaye. Chair, I thought we were going to be 20 following up on the past presentation on Chapter 3, to 21 close up some --
22 (Simultaneous speaking.)
23 CHAIR KIRCHNER: Oh, yes. Yes, you're 24 correct. I misspoke. Go ahead. Thank you.
25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
11 MEMBER HALNON: Right. We have just a 1
quick follow-up on Chapter 3. The Chapter 3 on 2
February 4th will be the 3.7, 3.8, and 3.9.2 aspects 3
of it. So it's a little confusing, the fact that we 4
split Chapter 3 up in a couple different places.
5 That's the clarification.
6 Tom, it's to you now, I believe.
7 MR. GRIFFITH: Yeah, good morning. Thomas 8
Griffith, Licensing Manager, NuScale Power. I wanted 9
to thank you for the opportunity to present on Chapter 10 16, the Loss-of-Coolant Accident Evaluation Model, and 11 for the opportunity to present an update for the high-12 impact technical issues.
13 I would like to recognize the efforts by 14 both the NRC staff and NuScale staff, and express my 15 appreciation of the efforts that have gone into the 16 review thus far. Furthermore, I would like to thank 17 my NuScale counterparts for supporting such an early 18 meeting. It is much appreciated for you all to be on 19 the phone at such an early time, but it is essential 20 that we have the opportunity to adequately discuss the 21 topics that we're going to talk about today.
22 So, in conclusion, thank you for this 23 opportunity, and thank you to the NRC staff, as well 24 as NuScale.
25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
12 MEMBER HALNON: Okay. Do you have --
1 who's presenting at this point?
2 MR. GRIFFITH: I have Gary Becker on the 3
line.
4 MEMBER HALNON: Okay. Gary, you're up.
5 MR. BECKER: Good morning. Make sure I 6
can be heard over there.
7 MEMBER HALNON: Yes. Give your name so 8
the court reporter, too.
9 MR. BECKER: Thank you. Gary Becker. I 10 am NuScale's Senior Regulatory Affairs Counsel. I've 11 been with NuScale for almost 15 years, where I serve 12 as our nuclear and licensing attorney. And, Walt, I 13 can report that it is, indeed, 5:30 in the morning 14 here, but unfortunately the moon is obscured, so it's 15 just very dark.
16 I appreciate the opportunity to -- before 17 you get into the heart of today's meeting - the 18 opportunity to discuss the Subcommittee's draft 19 Chapter 3 memorandum on the topic of standardization 20 and downstream licensing reviews. We don't have any 21 specific concerns with the memo's content, but I did 22 want to take a few minutes to share our views on the 23 topic and put those on the record.
24 NuScale recognizes the Subcommittee's 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
13 concerns for standardization and efficient licensing 1
reviews. We agree. We were concerned with, and paid 2
a lot of attention to, achieving standardization in 3
our SDA application. We applied lessons learned from 4
prior design reviews, up to and including the NuScale 5
DCA, and from recent reactor construction projects.
6 We drew upon the extensive experience of our 7
engineering licensing personnel in those efforts.
8 Just a high level on our application 9
approach, the NuScale design philosophy ensures 10 standardization by consolidating as much of the 11 safety-related SSCs as practical into the factory-12 fabricated NuScale power modules and the standardized 13 reactor building, which will be constructed using 14 modular construction.
15 Our SDAA follows a graded approach to 16 design information, providing detailed descriptions in 17 areas that are more important to safety and less 18 information in areas of lesser safety significance.
19 Reducing the bulk of our application in some areas 20 optimizes the user's and NRC's attention on what is 21 important versus what isn't.
22 As just one example, because our ultimate 23 heat sink is inside of a protected seismic Category 1 24 structure, it doesn't really matter, for the sake of 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
14 safe shutdown considerations, if a plant is water-1 cooled or air-cooled, next to a lake or in a desert.
2 So we have left some of those specific design details 3
to a downstream user.
4 The Subcommittee's Chapter 3
draft 5
memorandum discusses some COL items in Chapter 3 as 6
examples with respect to standardization. The design 7
information provided in the SDAA prescribes a complete 8
and final design, essentially complete and final 9
design, that must be built by a licensed applicant 10 referencing it, with limited and targeted allowances.
11 One of those allowances:
where final design 12 information is not yet known or is not included in the 13 FSAR description.
14 So, generally speaking, these are cases 15 where it's not practical for NuScale to complete the 16 design at this stage, or it's commercially prudent to 17 provide some flexibility for the applicant. In such 18 cases, the SDAA prescribes the safety implications, if 19 any, that must be considered and completed in the 20 detailed design.
21 Another allowance is where a license 22 applicant changes an aspect of the design that is 23 described in the FSAR, what we refer to as a 24 departure. Where departures are anticipated, the SDAA 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
15 prescribes confirmatory considerations in safety-1 significant areas. In all instances, the general 2
requirements to address safety considerations for 3
departures will apply.
4 I wanted to comment on -- note those 5
areas, because these allowances and processes are 6
central to making the standard design usable. While 7
a high degree of standardization is important to 8
realize the safety benefits intended by the 9
Commission, and standardization is essential to 10 achieve efficient plant licensing reviews, as you've 11 noted, we view that undue stringency and rigidity in 12 standardization would likely bring about the opposite 13 result. That is, if standardization processes are too 14 burdensome for future design applicants to use, they 15 won't.
16 As the Subcommittee's draft memo observes, 17 we expect a first-of-a-kind reference COL to complete 18 any necessary design details and address site-specific 19 features, and then commercial considerations will 20 drive that design to be repeated to the extent 21 possible in subsequent COLs. So these COL mechanisms 22 compliment the SDAA to ensure standardized, as-built 23 designs.
24 One comment on the topic of efficient 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
16 downstream licensing reviews. As noted in the 1
Subcommittee's draft memo, efficient, nth-of-a-kind 2
licensing reviews are essential to meeting the ADVANCE 3
Act objectives, and processes to focus review on 4
safety-significant issues would help in that regard.
5 For example, I discussed a moment ago the 6
potential for COLA departures. If a departure doesn't 7
alter the methodology or conclusions of the referenced 8
design, then it should be dispositioned quickly in the 9
staff and ACRS reviews. The design certification 10 rules provide a 50.59-like process for that purpose, 11 but NuScale has noted the SDA regulations do not 12 provide something similar.
13 We've urged in a ongoing rulemaking to 14 consider this issue, for the NRC staff to consider 15 this issue, and that's the Alignment of Licensing 16 Processes and Lessons Learned From New Reactor 17 Licensing rulemaking, addressed by SECY-22-0052, and 18 recently approved to go forward in a Commission Staff 19 Requirements Memorandum.
20 I note that the staff's position in their 21 draft-proposed rule falls short of our recommendation, 22 and we look forward to weighing in on that when the 23 proposed rule is published for comment. I bring it up 24 because that might be -- that rulemaking might provide 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
17 an area for efficient nth-of-a-kind licensing reviews 1
to be considered and approved, and might warrant 2
attention.
3 So, in conclusion, the NuScale SDAA 4
provides a highly standardized, optimized licensing 5
basis for future license applicants to follow. We 6
agree the reference in subsequent COL processes are an 7
important aspect of turning the standard design into 8
standardized plants, and in ensuring timely and 9
efficient licensing reviews in the process.
10 Allowances for limited variations in site-specific 11 designs are essential to using standardized designs, 12 and commercial incentives will minimize, and licensing 13 controls will address, the safety of such variations.
14 Importantly, our SDAA, and the standard 15 design processes more generally, would not allow the 16 degree of design divergence that occurred in earlier 17 standardization efforts under Part 50.
18 And, finally, we agree that an efficient 19 NRC review process for license applicants will be 20 important for the industry and the agency to meet the 21 objectives of the ADVANCE Act, while supporting our 22 objectives for efficiently building safe nuclear 23 facilities.
24 Thank you for the opportunity, and happy 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
18 to answer any questions that -- if you have them.
1 MEMBER HALNON: Any questions from members 2
in the room?
3 (No response.)
4 MEMBER HALNON: Online, any questions?
5 (No response.)
6 MEMBER HALNON: Gary, thank you. That was 7
an excellent summary and conclusion. This is -- I'm 8
sorry, James, this is Greg Halnon. Since I was the 9
primary author of the memo, we'll be going through the 10 memo in the next, I believe, full Committee meeting, 11 during our P&P session, our Practices and Procedures 12 session, so we'll get it fine-tuned there.
13 But I think the learning process that 14 we're going through, both in this one and what we did 15 in Kairos, and probably in the next couple, on what an 16 NOAK, or nth-of-a-kind, looks like, will be a key 17 input to the rulemaking coming up. So, I appreciate 18 that plug there, Gary. We think that's going to be 19 important, and incumbent on us as an agency to define 20 and understand how that's going to impact our 21 resources going forward so we can meet the ADVANCE 22 Act.
23 One last
- chance, any questions or 24 comments?
25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
19 (No response.)
1 MEMBER HALNON: Okay, Tom, it sounds like 2
-- Walt, if you're okay, we'll proceed on with Chapter 3
16.
4 CHAIR KIRCHNER: Yes, go ahead, Greg.
5 MR. GRIFFITH: All right. This is again 6
Thomas Griffith. Appreciate the opportunity to 7
present Chapter 16. We will have Karl Gross 8
presenting Chapter 16 in lieu of Gene Eckholt. And so 9
at this point I'll turn over to Karl Gross to begin 10 the presentation of Chapter 16.
11 MR. GROSS: Thank you, Tom. My name is 12 Karl Gross. I'm with NuScale. I joined them back in 13 2014, originally, primarily working on the technical 14 specifications. I'll be covering the tech specs and 15 the changes that have occurred to those tech specs 16 since the DCA was submitted.
17 Can I have the next slide, please? And we 18 can go -- do that again. There we go. That's the DOE 19 acknowledgment, of course, supporting our work. Next 20 slide. There we go.
21 NuScale included a
Part 4
Generic 22 Technical Specifications with the SDAA, consistent 23 with the statements of consideration for the 2007 rule 24 change to 10 CFR 52, where the Commission expressed 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
20 its expectation for the contents of the application 1
for design approvals to contain, essentially, the same 2
technical information required by a DCA.
3 We recognize that it's difficult to 4
perform a review of a design without having 5
accompanying technical specifications. Subpart E of 6
10 CFR 52 does not require technical specifications 7
for consideration, but, as I said, we went ahead and 8
included them.
9 We can go to the next slide, please.
10 These are several -- there were several changes that 11 resulted in changes to the GTS. Obviously, the most 12 noteworthy driver was the increase in thermal power 13 from 160 to 250. We'll cover the rest of these as we 14 go through the slides.
15 Next slide, please. Development of the 16 SDAA GTS, as we mentioned, started with the DCA model 17 tech specs as a model, and then we incorporated 18 changes resulting from the design and analyses using 19 criterion 50.36, applying those criteria. We also 20 considered industry Travelers, used the writer's 21 guide, the industry writer's guide, for format and 22 content, as appropriate, with, obviously, some little 23 differences there, since that was written for the 24 large PWRs and BWRs. And provided a summary of 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
21 discussions of this process in our technical report 1
which was submitted to the staff. That tech report 2
highlights the differences between the DCA and the 3
US460 GTS at the time of the SDAA submittal.
4 Next slide, please. The next couple 5
slides highlight some of the differences and changes 6
that occurred between the two sets of technical 7
specification GTS. One of the important ones on this 8
one I'd like to highlight is the remote shutdown 9
station LCO was removed from the tech specs because of 10 our passive design -- did a more careful review of it, 11 the way we implement it.
12 And next slide, some more changes. A 13 couple things to note here, one of them -- and most of 14 these will be addressed in other discussions, 15 obviously, the design itself or the analyses. But we 16 reduced the number of reactor vent valves, those are 17 the valves above the -- up high on the reactor vessel 18 from three to two. We also added an LCO, or what we 19 call ECCS Supplemental Boron System, which is 20 important for extended cooling periods to maintain 21 shutdown margin -- maintain a reactor shutdown.
22 Next slide, please. The NRC review of the 23 Chapter 16 and the proposed GTS resulted in 68 audit 24 items, which have all been resolved. There are no 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
22 RAIs outstanding specifically related to the tech 1
specs, however, other changes are driving, or have a 2
potential to drive, additional changes. So, as other 3
RAIs are resolved, we're watching those carefully.
4 MEMBER SUNSERI: This is Matt Sunseri. I 5
have a question about the review process.
6 MR. GROSS: Go ahead.
7 MEMBER SUNSERI: I know you're referring 8
to the NRC review there. You know, when I look at the 9
topical -- or the technical report and how they 10 construct these tech specs, you know, you outlined 11 clearly how you followed those. And I think, you 12
- know, in many cases this individual limiting 13 conditions for operation, or whatever, are derived 14 from technical, specific, you know, ways that the 15 system functions. But some of them you have to put 16 more thought into the situation or the circumstances 17 or the operating parameters or, you know, what can or 18 cannot happen at the plant. So it requires some 19 careful thought on, you know, thinking about the 20 scenario and then what would be bounding for including 21 in a technical specification.
22 So, it looks like you did a really good 23 job on that. But my question is, did you have any 24 kind of independent peer review to validate that you 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
23 didn't miss anything? Because you took out a lot of 1
stuff, based on your experience. You added a few 2
things, but, you know, it's the devil's in -- I don't 3
want to say that -- but, you know, sometimes it can be 4
hard to not know what you do not know from your 5
internal perspective, so having that peer, independent 6
peer, can open your eyes for things. Did you have a 7
peer review or anything?
8 MR. GROSS: Not so much independent, but 9
we do work very closely with our operating group and 10 also with our safety analysis group. They spend 11 extensive time, I'll say it this way, picking apart 12 our proposed tech specs. As you know from working 13 with operators, probably, versus licensing, we spend 14 a lot of time working with them to make sure that our 15 intent is clear and what the -- that it aligns with 16 the safety analyses. So, that was carefully 17 considered and we spent untold hours doing that.
18 It was -- the 50.36 identification 19 process, as described in the DCA tech report, was 20 basically reapplied for the SDAA. So there was also 21 a, I won't call it a clean sheet, but it was close to 22 a clean sheet review during the development of these 23 tech specs.
24 MEMBER SUNSERI: Okay, that's good. Just 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
24 a short follow-up, then. A lot of -- not a lot of 1
times, but it is possible that when you go into 2
operation you get similar scrutiny like this, so 3
should you find, when the plant is in operation, that 4
you have a tech spec that doesn't quite cover you --
5 I don't want to use the word, inadequate tech spec, 6
like it's commonly used, but, you know, you find 7
something that you might have missed that you need to 8
include, how would that be dealt with?
9 MR. GROSS: Depending on when it's 10 identified, you know, if we were actually at the 11 operating, where we've got fuel loaded -- in either 12 case, we're going to have to go through NRC approval, 13 obviously, to get a change identified.
14 I, personally, happen to have been through 15 that. And you're right, there's a -- there will be 16 changes. I hate to say it that way, but that's my 17 experience, anyway, every plant I've ever heard of, 18 and clarifications. Sometimes they can be made at the 19 basis level, which we can do ourselves, under the 20 rules. But if a change to the actual technical 21 specifications, we'll have to get Commission approval 22 beforehand. There'll be a continual feedback loop 23 during that process.
24 MEMBER SUNSERI: Okay, perfect. Thank 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
25 you.
1 MEMBER KIRCHNER: Karl, this is Walt 2
Kirchner. The third sub-bullet on the slide that's in 3
front of us right now under "noteworthy changes,"
4 maybe this is a question for the closed session, but 5
what caught my eye was that the geometric form of the 6
boron pellets, could you elaborate on that?
7 I am presuming that just is a time, 8
depending on the form of the pellets that would 9
perhaps impact the time that it would take for that to 10 dissolve and go into solution, is that what you are 11 referencing here?
12 MR. GROSS: That's basically what it 13 amounts to is the dissolution rate that was assumed in 14 the safety analysis.
15 MEMBER KIRCHNER: Okay.
16 MR. GROSS: Our safety analysis people 17 maybe can jump in if they feel they want to.
18 MEMBER KIRCHNER: But it manifested itself 19 in the form of a time specification?
20 MR. GROSS: Actually I think we modeled it 21 based on an expected and demonstrable dissolution 22 rate.
23 MEMBER KIRCHNER: Right.
24 MR. GROSS: But we didn't, you know -- it 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
26 will be in the COLA is where we are including that 1
requirement, the actual specifics.
2 MEMBER KIRCHNER: Okay.
3 MR. GROSS: Right now it's kind of like, 4
you all remember the baskets in the bottom 5
containments and there was issues early on about them 6
turning into a big slug, I guess is a nice way of 7
saying it.
8 MEMBER KIRCHNER: Right.
9 MR. GROSS: We are trying to avoid that 10 and we have avoided it by including requirements for 11 the form to ensure that they dissolve in an 12 appropriate timeframe.
13 MEMBER KIRCHNER: Okay, all right. Thank 14 you.
15 MR. GROSS: Mm-hmm.
16 DR. SCHULTZ: Karl, this is Steve Schultz.
17 Your last bullet here, the change that you had to make 18 associated with the generic tech specs with regard to 19 the steam generator program, could you expand on that, 20 what was it that you had to address there?
21 MR. GROSS: I'll defer some of this to 22 others, but it basically boiled down to the Staff had 23 concerns with the initial inspection to ensure that we 24 had adequate inspections during the early operational 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
27 days. I don't know --
1 MR. SNUGGERUD: This is Ross Snuggerud 2
with NuScale.
3 MR. GROSS: Thank you, Ross.
4 MR. SNUGGERUD: Yeah. What Karl said is 5
basically correct. There was a desire to make sure 6
that in support of our position on how the steam 7
generators would function and how the steam generator 8
would potentially be impacted by density wave 9
oscillations at low power, the industry's behavior of 10 spacing out the steam generator inspections is more 11 implied than stated, and so in our generic tech specs 12 it will state that we need to do certain fractions of 13 the inspections every refueling outage.
14 So, even though that's the way the 15 industry does it generically, it's implied in the 16 industry and its industry behavior. For NuScale, they 17 wanted to ensure that we actually have our owners 18 doing partial inspections following the first and 19 subsequent and not just waiting for the full 20 inspection period to do the entire steam generator 21 inspection.
22 MS. BLUMSACK: This is Erin Blumsack from 23 NuScale. I just wanted to clarify what Ross said. We 24 have a COL item in Chapter 5 indicating that we will 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
28 do inspections in a staggered basis until full 1
inspection of the tubes is complete.
2 In the technical specifications we didn't 3
change that, we reduced the amount of time between 100 4
percent tube inspections and we also bracketed the 5
tube plugging criterion.
6 DR. SCHULTZ: Thank you.
7 MEMBER BIER: Quick question from Vicki 8
Bier. Can you talk about whether the staggered steam 9
generator inspections have adverse cost implications 10 for users or do you think the cost is about 11 equivalent?
12 MS. BLUMSACK: This is Erin Blumsack from 13 NuScale. We haven't evaluated the potential cost, 14 however, the steam generator program indicates that we 15 will complete a degradation assessment and that will 16 inform future inspections.
17 So we're not expecting that anyone would 18 do anything they --
19 (Audio interference.)
20 MEMBER BIER: Thank you.
21 MEMBER MARTIN: This is Bob Martin. I had 22 a question, and I was planning on using it for the 23 LOCA but I'll bring it in here, are there tech specs 24
-- Someone's got a hot mic.
25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
29 Anyway, are there tech specs associated 1
with leak rates of the RVVs and RRVs and then the 2
other part of the question that would be a stretch is 3
how do you mitigate excessive leaks, leak rate?
4 MR. GROSS: This is Karl Gross again. The 5
RVV and RRV leak rates would be the same as those for 6
the RCS as a whole. So, yes, that is addressed in 7
Section 3.4 of the tech specs.
8 As far as mitigation, as you know probably 9
the containment is operated under vacuum conditions 10 and we have systems to maintain it there and monitor 11 the pressure there and remove leakage --
12 (Simultaneous speaking.)
13 MEMBER MARTIN: Okay.
14 MR. GROSS: Okay. So I guess those --
15 MEMBER MARTIN: Is that also for say 16 possible air ingress? You have a vacuum, so 17 presumably you have a line to, you know, vacate it.
18 MR. GROSS: Yeah.
19 MEMBER MARTIN: Same kind of thing, you 20 monitor for potential air ingress into the 21 containment?
22 MR. GROSS: Yeah, not specifically as air 23 but we watch the pressure, I believe, unless something 24 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
30 MEMBER MARTIN: Okay. That would be the 1
same thing.
2 MR. GROSS: Yeah.
3 MEMBER MARTIN: Yeah.
4 MR. GROSS: Next slide, I guess, if we're 5
ready. One change that has occurred as we determined 6
it appropriate to include an LCO for the passive auto 7
re-combiner, that it will be discussed further in the 8
Chapter 6 presentation.
9 I know that's coming up in the future as 10 to why it was included. I want to point that out 11 there is a new LCO for that.
12 Next slide. I think that's it. Thank 13 you, unless anybody has any more questions.
14 MEMBER SUNSERI: This is Matt Sunseri.
15 I've got maybe one or two more. Excuse me. I noticed 16 that the introduction to the technical report stated 17 that the tech specs are for a single module and I 18 understand that.
19 So I presume that each module would then 20 have its own technical specifications, would that be 21 accurate?
22 MR. GROSS: That is correct. The intent 23 is to keep them aligned as much as possible. I 24 recognize I think over the life of a facility there 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
31 may be some small variations as changes are 1
incorporated into the design.
2 If they are, that's obviously not our 3
intent, but, yeah, there would be a complete set for 4
each module.
5 MEMBER SUNSERI: Yeah. So -- yeah, well 6
that's where my question, and I wouldn't call it, 7
maybe concern is too strong of a word, but at least my 8
experience with multiple reactor sites, I'll call it 9
that, that, you know, in some cases you have a common 10 control room but they are segregated enough such that, 11 you know, you have Unit 1 operators, you have Unit 2 12 operators, you have Unit 1 tech specs, you have Unit 13 2 tech specs, whatever, and they are pretty separate 14 and the operating crew can, you know, pretty much 15 guarantee they know which unit they're on and which 16 set of tech specs they're in, but the way your control 17 room is going to be operated you're going to a few 18 operators covering as many as six modules with six 19 different tech, potentially six slightly different 20 tech specs. That seems like a pretty challenging 21 human performance challenge.
22 What are your thoughts on how you're going 23 to manage that and is there anything that needs to be 24 put back into the tech specs, into a single unit tech 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
32 spec. Single module for you guys.
1 MR. GROSS: Right now we're obviously 2
we're not aware of anything that would be required to 3
distinguish between them. However, recognizing that 4
there's a potential for that in the future, we're 5
going to leverage heavily I think our automated 6
systems.
7 As you know the control room is very, can 8
be automated, almost all the functions. It's kind of 9
spooky if you're used to old plants. But, yeah, so 10 hopefully any changes we can address that way or 11 incorporate appropriate controls to ensure alignment 12 that way to support our operators.
13 Ross, would you like to jump in on any of 14 this?
15 MR. SNUGGERUD: I mean I think the most 16 likely condition and the one we have talked to our 17 potential operators about is the fact that we are 18 going to be implementing some change on a module and 19 that's going to require us to cycle through the tech 20 specs, which is one of the advantages about having 21 individual tech specs.
22 If we need to make a modification and then 23 we're implementing that modification systematically 24 throughout the modules over a course of refueling 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
33 outages we can keep the tech specs up to date for the 1
individual modules that have had that configuration 2
change.
3 We don't really expect a lot of module-4 specific technical specification changes, so, you 5
know, at most I think we would have two versions of 6
the tech specs at any given time.
7 But, you know, as Karl suggested, we are 8
going to be providing the operators with lots of 9
support through the interface and there are lots of 10 different ways to address that.
11 Certainly it's not -- you know, your 12 concern is valid if we started to have deviations 13 between the tech specs, but the idea of keeping the 14 tech specs individual to the modules was intentional 15 and largely to support operations at the plant.
16 MEMBER SUNSERI: Yeah, that's good and I 17 appreciate that. You know, having been in these kind 18 of control rooms, or situations before, you might just 19 give some consideration to how it would flag or 20 highlight, you know, any potential differences so that 21 you guys -- human performance, right, you know how 22 that goes.
23 MR. SNUGGERUD: Yeah, absolutely, we 24 agree.
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34 MEMBER SUNSERI: Yeah. Hey, and I have 1
one more technical question that I'm not sure it's 2
going to come up in an individual chapter review so 3
I'm going to ask it now.
4 It deals with the definition of MODE 2 and 5
MODE 3, hot shutdown and safe shutdown and there is a 6
various combination of passive cooling or not passive 7
- cooling, minimal temperature for criticality, 8
whatever.
9 I understand all that, but I just want to 10 ask one question. Is there a reactivity -- are both 11 those conditions limited to less than 0.99 or, yeah, 12 0.99 k-effective?
13 MR. GROSS: That is correct.
14 MEMBER SUNSERI: Okay. It was -- I 15 thought it was when I read the GTS, but, you know, 16 some of the documentation I read it wasn't clear that 17 that condition was always -- so it is truly less than 18 0.99, okay. I'm good. Thank you.
19 MR. GROSS: Yeah. Yeah, it's specified 20 that way in the definition of MODES, that table at the 21 end of Section 1.1 of the tech specs.
22 MEMBER SUNSERI: Yeah, I saw that but then 23 I also know that those aren't the real tech specs, 24 right, that's just a generic one that you submit for 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
35 reference?
1 MR. GROSS: Yes. But the intent is that 2
that's where it will be.
3 MEMBER SUNSERI: Okay, all right.
4 MR. GROSS: The MODE tables are pretty 5
well defined.
6 MEMBER SUNSERI: Okay. That's all I had.
7 Thank you.
8 MEMBER KIRCHNER: Members, any further 9
questions of NuScale before we turn to the staff?
10 (No response.)
11 MEMBER KIRCHNER: Mike, correct me if I've 12 got the agenda wrong. Are we going to complete 16?
13 MR. SNODDERLY: No, you're right. You're 14 right, Walt. We found that --
15 MEMBER KIRCHNER: Okay.
16 MR. SNODDERLY: Unless the Committee gives 17 us other feedback, but we find it's best to complete 18 a chapter and then the staff going --
19 (Simultaneous speaking.)
20 MEMBER KIRCHNER: Right. Right. So if 21 there is no further questions I think we're ready to 22 move to the Staff's presentation.
23 MEMBER HALNON: Okay. Give us a few 24 minutes to change the seats and computers and I'll let 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
36 you know when we're ready, Walt.
1 MEMBER KIRCHNER: Okay, mm-hmm.
2 (Pause.)
3 MEMBER HALNON: Okay, Walt, we're going to 4
get started again.
5 MS. SCHILLER: My name is Alina Schiller.
6 MEMBER HALNON:
Please turn your 7
microphone on. Speak right into it.
8 MS. SCHILLER: Good morning.
9 MEMBER HALNON: That's not close enough.
10 MS. SCHILLER: Good morning.
11 MEMBER HALNON: That's better. Thank you.
12 MS. SCHILLER: My name is Alina Schiller.
13 I am a project manager with the Office of Nuclear 14 Reactor Regulation, Division of New and Renewed 15 Licensees, New Reactor Licensing Branch.
16 I
would like to thank the ACRS 17 Subcommittee, NuScale Power, LLC, and the general 18 public for entertaining the NRC for the presentation 19 of the Staff's safety evaluation of NuScale Standard 20 Design Approval Application from Chapter 16 and the 21 Part 2, Revision 1.
22 NuScale 70 Part 2, the Final Safety 23 Analysis Report, Chapter 16, Technical Specifications 24 and Part 4, Generic Technical Specifications, Revision 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
37 0, in December 2022 and Revision 1 in October 2023.
1 From March 2023 to August 2024 the NRC conducted a 2
regulatory audit of FSAR Chapter 16 and Part 4 which 3
generated 68 audit issues. All audit issues were 4
resolved in the audit.
5 Fifty-two audit issues resulted in NuScale 6
submitting supplemental information to address 7
questions raised during the audit. No requests for 8
additional information were issued in Chapter 16.
9 We are here today to discuss the Staff's 10 advanced safety evaluation of Chapter 16 and Part 4.
11 The contributors were Craig Harbuck, the lead 12 technical reviewer and today's presenter, supported by 13 Steve Smith, Clint Ashley, Josh Wilson, all with the 14 Technical Specifications Branch, and the project 15 manager for Chapter 16 and Part 4 supported by 16 Getachew Tesfaye, the lead PM for NuScale SDAA.
17 This slide lists the sections in the FSAR 18 Chapter 16 and Part 4. I am turning it over to the 19 NRC subject matter expert, Craig Harbuck.
20 MR. HARBUCK: Good morning.
21 MEMBER HALNON: Craig, you're going to 22 have to move that real close, as close as you can move 23 it to yourself comfortably.
24 MR. HARBUCK: How's that?
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38 MEMBER HALNON: That's good if you keep it 1
that way.
2 MR. HARBUCK: Okay. So I am going to go 3
through changes that we thought were worth mentioning 4
outlining the differences between the DCA certified 5
generic tech specs, the US600 design, and then 6
reviewed the SDA.
7 This slide is to point out one of the key 8
differences in the definition of MODE 3. It was 9
adopted in the SDA. Essentially what has happened is 10 that the MODE 3 in the DCA began at the minimum 11 temperature for criticality, but the MODES 1 and 2 12 covered temperatures above that.
13 There is one thing I would like to point 14 out that might not be obvious to everyone about the 15 MODE 1 definition. It actually corresponds to MODES 16 1 and 2 that you see in normal PWR plant for the 17 operating fleet.
18 So as was mentioned earlier in NuScale's 19 presentation, you can enter MODE 3 from 1 or 2 by 20 initiating passive cooling without first cooling down 21 the module to below the, well the minimum temperature 22 for criticality.
23 That particular temperature which that is 24 was lowered in the SDA from the DCA value of 420 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
39 Fahrenheit to 3.5 and the way you initiate passive 1
cooling when you are at the normal operating 2
temperatures would be to either open up the ECCS 3
valves or by initiating the decay heat removal system.
4 Both of these would result in a
5 significant transient and a fairly quick cool down and 6
so if you're in an action statement and it says shut 7
the plant down to MODE 3 and a lot of times it will 8
indicate, it may give a caveat on a lower temperature 9
range than MODE 3, but typically you wouldn't 10 implement that action statement by initiating passive 11 cooling.
12 The preferred method would just be to use 13 secondary heat sink systems to do that.
14 MEMBER HALNON: So, Craig, this is Greg.
15 What was the purpose then, I mean if the normal 16 shutdown is going to be 48 -- or whatever, a k-17 effective less than 0.95 and you cool it down to less 18 than 345 is that in there for an operational transient 19 situation or reactor trip of like a loss of power or 20 something?
21 MR. HARBUCK: I can speculate about what 22 the operational conditions might be where this would 23 be appropriate, but --
24 (Simultaneous speaking.)
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40 MR.
HARBUCK:
There must be some 1
operational advantage to being able to say that you're 2
in safe shutdown until you run actually your MODE 2 3
and not have to move that MODE change as long as 4
you're being passively cooled, but I would have to 5
refer to NuScale to address that question.
6 MEMBER HALNON: Okay, that's fine. It's 7
not a game changer, it was just more of curiosity. Go 8
ahead.
9 MR. HARBUCK: Okay. But it is the cycle 10 that the nature of the design and how to operate it 11 because you do have to think about these things when 12 you are deciding what's the appropriate way to account 13 these information constraints from the definitions.
14 I also want to point out that the MODE 3 15 definition before you get to -- well there is a 16 footnote on some instrumentation functions which tell 17 you that certain reactor trip and ECCS functions or 18 PSF functions don't have to be operable if you have 19 just one control rod mechanism being energized in any 20 fast term is being capable of withdrawal and typically 21 that particular footnote is stated as any (audio 22 interference) fairly unique NuScale design features 23 related to coupling and uncoupling control rods.
24 They changed the footnote so that you 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
41 could have one rod that you were trying to couple or 1
uncouple would not by energizing the control rod drive 2
to manipulate the rod to do that would necessitate 3
that then goes, is corresponding to instrumentation 4
functions probabilities and that's one difference from 5
the rule, that normal caveat in the footnote in 6
operating.
7 Then going to the next slide, there was 8
one that was adopted recently, recently approved, 9
relatively recently, that clarified the definition of 10 pressure boundary leakage and NuScale elected to adopt 11 those changes.
12 So this mark up here shows the change from 13 what we had in the DCA and what we have now and this 14 is consistent with the industry's understanding of 15
- how, and the Staff's understanding of
- what, 16 particularly that last added sentence.
17 It's something that actually in the bases 18 for the LCO related to leakage that is simply up and 19 made part of that definition.
20 Next slide. And going to the safety 21 amendments chapter, this shows you that the comparison 22 of the different correlations from the DCA to the SDA 23 and the NSP4 correlation was maintained but the other 24 ones were not and then for certain operational 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
42 aspects, which NuScale could elaborate on, there was 1
an additional correlation that was added on.
2 The other reactor safety limit on peak 3
center line temperature, the fuel did not change, 4
that's quoted there as saying that, and then because 5
of the higher operating pressure in the higher powered 6
design we moved from 1850 to 2000 psi.
7 So the safety limit under, on the reactor 8
coolant system pressure has increased as noted there 9
on the slide.
10 DR. BLEY: This is Dennis Bley. Could you 11 back up a slide to Number 6? Thank you. No. There 12 we are.
13 MR. HARBUCK: Yes.
14 DR. BLEY: Leakage past seals piping 15 gaskets is defined here as not a pressure boundary 16 leak and I don't know that you have specified 17 equipment to the level that would allow us to know how 18 big such leaks could be.
19 I've seen some pretty big ones, like if 20 seals blow out or something like that. Why do you 21 express it this way rather than in terms of maybe a 22 pounds per hour or something, you know, a quantitative 23 definition of what you mean by pressure boundary leak.
24 MR. HARBUCK: The way your point is 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
43 addressed is in the LCO for a leakage for the reactor 1
coolant system --
2 DR. BLEY: And --
3 MR. HARBUCK: And so there is no pressure 4
boundary leakage, that's not allowed. If you identify 5
that then you would be on a shutdown track.
6 That brings up another point, is that 7
currently NuScale in their containment evacuation 8
system it's just what maintains the vacuum in the 9
containment of their operation.
10 They don't currently have a way of 11 determining any leakage that is collected about what 12 the source of that leakage is in terms of is it 13 pressure boundary leakage or is leakage past some, 14 like this phrase says, or is it, you know, is it 15 coming from a secondary system, is it coming from a 16 leaky external system like that containment flood and 17 drain system, or is it, let's see, is it coming from 18 the CVCS, or is it coming from like a feedwater line 19 or something, which is not part of the pressure 20 boundary within the CF.
21 So they essentially will treat any leakage 22 that's detected if I understand correctly, and correct 23 me if I'm wrong, but I think it would essentially 24 treat any leakage they collected in their containment 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
44 evacuation system as unidentified leakage, and there 1
is a limitation on that in the LCO.
2 MEMBER HARRINGTON: Dennis, this is Craig 3
Harrington. That added sentence is consistent with 4
the ASME code treatment of pressure boundary that 5
packing seals, gaskets, that's not pressure boundary.
6 DR. BLEY: Okay.
7 MEMBER HARRINGTON: It is, but --
8 DR. BLEY: They are being consistent. I 9
understand, but they are being consistent with the 10 standard. Okay. You know, it just felt funny to me.
11 Thanks, Craig.
12 MR. HARBUCK: Okay, now we can go to the 13 next slide. I think we've covered this, yeah, so 14 let's go to Chapter 3 now. The remaining slides will 15 roughly focus on Chapter 3.
16 So earlier it was alluded to that some 17 specifications were, some LCOs were removed that would 18 have been an additional DCA to your tech specs and 19 then the renewal LCO had it in the SCA and this 20 provides a list of those.
21 The bullets under the two sections that 22 were removed continue to provide some, point to some 23 of the rationale for why those removed and why it is 24 acceptable to do that and I'll just briefly mention 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
45 those.
1 A remote shutdown station in a typical 2
operating plant would have the ability, would have 3
controls as well as indications that from which 4
personnel who had evacuated the control room be able 5
to shut down the plant and monitor its condition in a 6
safe shutdown situation.
7 In the DCA, the imagined remote shutdown 8
station was only going to have indication and no 9
controls per se because of the envisioned scenario was 10 that if there was a need to evacuate the control room, 11 part of doing that would be to shut down all the 12 reactors, and because of the design, where there 13 really is no operator action needed to ensure you are 14 in safe shutdown after that, then there did not seem 15 to be a need for any duplicate controls in a separate 16 station outside the control room.
17 The indication monitor, it was pointed out 18 during the review that the I&C equipment rooms 19 associated with each module will have, also have 20 indications that are explained in the control room and 21 you can find those in the same information and, 22 therefore, that was seen as being sufficient to 23 monitor the status of the modules in the event of a 24 control room evacuation.
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46 There was also -- and typically the 1
controls you need to make sure that you achieve safe 2
shutdown and operating plants would require, you would 3
have in the safety analysis you might have some 4
assumptions about actions that operators would take 5
and for those you have what's called Type A post-6 accident monitoring variables that would be seen as 7
needed to provide information the operators would need 8
in terms of to properly conduct that shutdown, but 9
there are no such, there are no Type A variables in 10 the NuScale design, which is another reason they don't 11 have a main LCO also.
12 The in-containment secondary piping 13 leakage in the DCA we address this idea to monitor the 14 leakage and you can use the leak-before-break method 15 to provide yourself assurance that you would be able 16 to recognize when you had the potential for a high 17 energy line break into avoiding any resulting pipe 18 movement that could damage any other equipment in 19 containment and you'd be able to shut down and address 20 that because there would be enough time to do that.
21 In the SDA they have determined that based 22 on the Staff guidance note on the BTP 3-4 and 23 industry's guidance, the standards, that address 24 certain pipes, size of pipes, and that sort of thing, 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
47 that they have this exclusion criteria and it was 1
determined that the high energy lines in containment 2
satisfy those criteria, so that LCO was deemed not 3
being necessary and the Staff has agreed to that.
4 So that's what we have removed or omitted, 5
but what's been added was, as they mentioned before, 6
the ECCS Supplemental Boron System and that's for 7
assuring that you have adequate shutdown margin for 8
long-term cooling.
9 That system is also implemented to make 10 sure that it's implemented after a reactor trip if you 11 are in a situation where a combination of xenon 12 transients and cool down could get you starting to 13 approach your re-criticality situation with water 14 flowing from the containment.
15 The idea was to make sure that the system 16 actuated and that's a passive design, it's hands off 17 for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. So after eight hours, unless the 18 operator has determined it's not necessary and they 19 can block the system from actuating, there is an 8-20 hour post-reactor trip actuation time on ECCS 21 primarily to be able to initiate the dissolving of ESB 22 pellets.
23 That timer is addressed, the verification 24 is addressed in the actuation logic LCO in SR-3333.
25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
48 The other, another LCO that was added that could very 1
well have been included in the DCA but there was, or 2
I guess NuScale recognized that during MODE 4 when you 3
are moving the module, the passage of decay heat 4
during that evolution is, requires there to be a 5
sufficient volume of water in the containment and so 6
to preclude also any inventory in the containment met 7
within diminished heat transfer capability. They 8
wanted to make sure that any isolation valves from the 9
containment to the outside would be closed.
10 So that is what this LCO is designed to 11 do. It doesn't have to meet the same leakage 12 requirements that you have for like containment 13 isolation valves in that LCO or to make containment 14 operable.
15 It's just designed to maintain the 16 critical inventory to ensure you have adequate cooling 17 and to what the module on the disassembly stand over 18 the refueling area and you've unseated the containment 19 off of its lower portion and, therefore, then 20 everything is filled up to whatever the pool level is.
21 Okay. Then it was also mentioned that the 22 passive auto catalytic be combined and it was 23 determined to satisfy Criterion 3. It was determined 24 to be a safety-related system and it was also somehow 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
49 justified as meeting this criterion in 50.44.
1 If anyone has a question about that I 2
would ask NuScale to respond as I am not that familiar 3
with what those very non-specific words in 50.44 4
correctly say. Any questions about that? Okay.
5 So next slide. So there were -- to me it 6
would be I guess one more significant change was an 7
instrumentation in terms of the instrumentation fluxes 8
that actuate the ECCS and also decay heat removal 9
system, which will be on the next slide.
10 But they changed the way they're measuring 11 level in the riser in the SDA over what they had in 12 the DCA and, therefore, they have ECCS initiation 13 primarily occurring on the riser level.
14 A low level which is designed to protect 15 and trying to prevent uncovery of the top of the riser 16 is blocked and if you go below 500 degrees it's 17 interlocked and it does that.
18 The low riser level is designed to actuate 19 ECCS before you uncover these holes that are in the 20 side of the riser down lower in the vicinity of the 21 steam generator in the downcomer to make sure that 22 there is adequate flow of the high, relatively high 23 concentration reactor coolant in the riser into the 24 downcomer region to alleviate any dilution effects 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
50 caused by evaporation that any of them had which ECCS 1
has now furnished. As was also mentioned --
2 (Simultaneous speaking.)
3 MEMBER MARTIN: Question real quick. This 4
is Bob.
5 MR. HARBUCK: Okay.
6 MEMBER MARTIN: And maybe this is a 7
NuScale question, but the level instrumentation is 8
that really a DP or is it --
9 MR. HARBUCK: That is done with a 10 specialized kind of discreet -- There are probes that 11 detect --
12 MEMBER MARTIN: Okay.
13 MR. HARBUCK: And those functions actually 14 have like a 60-second delay on when you actually get 15 a change in the signal.
16 MEMBER MARTIN: Okay.
17 MR. HARBUCK: So I think that's built into 18 the substance of the safety analysis. I don't know if 19 there is any software or other I&C magic going on that 20 would interpolate between those discrete levels 21 because they're not that far apart.
22 MEMBER MARTIN: Okay. That's just 23 something I wanted to clarify.
24 MR. HARBUCK: And that's another question 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
51 which will have, may warrant further investigation for 1
a COL applicant, but --
2 MEMBER MARTIN: Okay. All right, that 3
answers my question. Thank you.
4 MR. HARBUCK: Yes. Okay. Again, they 5
reduced the number of reactor vent valves from three 6
to two and I don't know if that was a change in the 7
size of the valve or if the third valve had simply 8
been there for some other redundancy reason, but to 9
accomplish the ECCS function you need only one reactor 10 vent valve and one recirculation valve.
11 So from that standpoint they are redundant 12 and then each valve itself is also doubly redundant 13 on, that's redundant itself. There are two actuation 14 solenoids on each valve that are separate to the 15 channels of the module protection system that have to 16 be de-energized because you reached the ECCS actuation 17 setting level for those valves to open.
18 Now there is one thing I want to -- this 19 was the subject that we were hitting on and I just 20 wanted to mention it.
21 For most events your reactor vent valves 22 continue to remain shut so that you don't introduce 23 the containment into the initial response to the event 24 and handle it like the decay heat removal system, a 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
52 reactor trip, what have you, then you may not need to 1
change anything with the containment.
2 And so to maintain those valves shut 3
during those kinds of transients and also during 4
normal operation, the EDAS to which is, that implies 5
special available or quality controls, you know, 6
augmented quality is what they call it.
7 So that -- because before the DC those 8
valves would not open until the pressure difference 9
between the containment and the RCS reached a certain 10 level, a certain level, thereby that had the effect of 11
-- even if you would -- if the actuation set point 12 until you reach that IAB, which was the more 13 mechanical kind of inhibiting of the opening of the 14 valves, the valves wouldn't open.
15 But that's been removed in the SDA with 16 the DES liability has been, combining the system and 17 all of that has been in these areas in which it needs 18 to be addressed or there was any kind of concern that 19 was resolved here.
20 But the reactor recirculation valve still 21 has an IAB and it's helping to still be delayed even 22 if you get to set point. So that's all I wanted to 23 say about that. Next slide.
24 MEMBER ROBERTS: Hey, Craig, it's Tom 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
53 Roberts. Can you clarify, you said there were two 1
solenoids in each of the vent valves and recirculation 2
valves?
3 MR. HARBUCK: That's correct.
4 MEMBER ROBERTS: Is there a tech spec that 5
they both have to be in service?
6 MR. HARBUCK: Yes.
7 MEMBER ROBERTS: Okay. So the intent is 8
that either one will trip a valve or either one will 9
heat it up?
10 MR. HARBUCK: You would need both of them 11 to actuate to cause the valve to open.
12 MEMBER ROBERTS: Okay. So from a tech 13 spec perspective let's say the safety position of the 14 valve would be open and so why would there be a tech 15 spec that says those solenoids have to be in service?
16 MR. HARBUCK: Well if the solenoid -- if 17 they are not in service they would de-energize and the 18 valves are going to open. This is a fail/safe design 19 on all valves to go to their safety position, so this 20 result was not, did not be -- I suppose you could 21 operate with one of them out of service.
22 As long as one of thems keeping it 23 closed, but it's not seen as being a safety-related 24 function. It's called the ECCS holds function. So 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
54 the -- we could put NuScale on to address this. They 1
would probably do a better job.
2 MEMBER ROBERTS: I guess we'll renew the 3
question later. It just seemed a bit odd. Certainly 4
it's prudent, right, definitely it's prudent if you 5
have the redundancy to have it because you don't want 6
to add inadvertent actuation, but I'm not sure why 7
it's a tech spec because the -- unless there is some 8
concern that inadvertent actuation is a safety 9
concern.
10 MR. HARBUCK: Well the tech spec is 11 focused on not having both of them for the purposes of 12 keeping the valve shut. It's there to make sure that 13 they really de-energize than actuate it.
14 So for tech spec purposes the valves would 15 be inoperable if for some reason you removed power and 16 they didn't change position and, therefore, the valve 17 could not, it would not open.
18 Those basically have a hydraulic lock on 19 the valve keeping it shut and then you de-energize the 20 solenoids the valves open and that removes the lock, 21 so it is kind of complicated in terms of the logic, 22 thinking about it. Anything more to add?
23 (No response.)
24 MR. HARBUCK: Okay. Next slide, decay 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
55 heat removal system. There is not much to say about 1
this other than that we added a
few more 2
instrumentation functions which would cause an 3
actuation of the decay heat removal system.
4 At the bottom there I list the three, 5
those three functions and the events that those 6
functions are designed to mitigate and they all boil 7
down to putting yourself in a safe shutdown situation.
8 That's when you set your MODE to 1 and you 9
would initiate a decay heat removal system that's 10 going to put you in MODE 3, and so that -- any other 11 questions about that?
12 (No response.)
13 MR. HARBUCK: I believe that's the last --
14 oh, yeah, there's one more slide. So we'll go to the 15 next slide. Oh, there's two more slides. Maybe this 16 is it. Yes.
17 Okay. Another change that happened, 18 another change that I want to mention was the boron 19 dilution control, LCO-319. There is a system in 20 NuScale that allows you to heat up the RCS called 21 module heat up system and the way the system works is 22 you have the heat exchanger that on one side is 23 supplied with steam from the non-safety source that 24 generated that, then you take the discharge out of the 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
56 CVCS and the discharge, I mean the injection line, and 1
the diverted flow by the heat exchanger and then it 2
comes back downstream of where it left back to the 3
injection line warmer and then the injection line 4
terminates in the riser section of the reactor vessel.
5 So in that way you can heat up the unit 6
since there are no reactor coolant pumps to do that 7
sort of thing or to heat up the coolant, but they 8
don't have a module heat up system heat exchanger for 9
each module.
10 There is a common one, so it was 11 recognized there was a potential that if you had 12 errors in alignment you could connect one CVCS system 13 from one module onto another module.
14 So this was added to the LCO and clarified 15 in the surveillance requirements to check the 16 alignment to make sure, you know, you never had more 17 than one module aligned with the module heat up 18 system.
19 That's usually for a relatively brief time 20 when you're starting up and you want to heat up the 21 system. So it's not a likely thing to occur, but it 22 was to see if that was a potential error that could be 23 addressed where safety could made operationally that 24 could be addressed by highlighting it in the LCO for 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
57 boron dilution control.
1 In the -- I think I mentioned earlier 2
about the MODE 3 applicability of selecting a reactor 3
trip in a de-mineralized water system isolation 4
instrument functions and that the footnote for those 5
functions for MODE 3 is incapable of withdrawal of 6
more than one control rod assembly.
7 So that was a difference from your 8
regular, the usual definition, and also a difference 9
from what they had in the DCA, because I think this 10 particular concern was not identified in the DCA or it 11 was not being needed to -- it just didn't come up.
12 The last thing, it was alluded to earlier 13 there were some changes from what was in the standard 14 tech spec and what was in the DCA relative to the 15 steam generator requirement and some items here that 16 were changed was the period between of having to 17 inspect all of the tubes after the initial -- I think 18 at the end of the first refueling outage there is, or 19 at the first refueling there was supposed to have been 20 another full inspection of all the tubes.
21 But subsequent to that you have 72 22 effective full power months in which I remember they 23 were going to do staggered in the sense of the valves 24 on the tubes until, so by the end of 72 effective full 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
58 power months you would have all your tubes expected to 1
be in.
2 So that's -- and previously had been like 3
96 months, so it was reduced. We have folks in the 4
room that maybe could address why that is. The other 5
parts of the description of the inspection discussion 6
in this program specification talked about the fact 7
that RCS pressures on the outside of the tubes and the 8
tubes are susceptible primarily to collapse from 9
collapse or buckling rather than bursting.
10 So that does allow for some other 11 differences that, you know, I think it's Chapter 4, 12 five, Chapter 5, where they go into more detail about 13 this.
14 Then this -- as we did in the DCA we put 15 a value of 40 percent for the criteria recognizing 16 that a COL applicant might have a different -- might 17 have had new information or new rationale and might 18 have a different number, so that's what the use of the 19 bracketed information is in the tech specs and in the 20 bases also. That falls under COL item 60.1-4.
21 I believe that's the last information 22 slide and this is a conclusion listing the regulations 23 that govern the tech specs that we have determined 24 that we are in compliance with those.
25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
59 MEMBER MARTIN: This is Bob Martin. Now 1
did you have RAIs --
2 MR. HARBUCK: We have no RAIs as far as 3
the process.
4 MEMBER MARTIN: Yeah.
5 MR. HARBUCK: I think we've managed to 6
resolve everything in the context of audit follow-up.
7 MEMBER MARTIN: Okay, all right. A pretty 8
straightforward review?
9 MR. HARBUCK: Yes, it was. A lot of our 10 issues were related to problems in other chapters, but 11 we always hear about it last. That concludes --
12 MEMBER HALNON: We have a NuScale person 13 with their hand up. Did you have a clarification, 14 Tyler?
15 MR. BECK: Hi. This is Tyler Beck with 16 NuScale. I just wanted to clarify about the ECCS trip 17 valve and the solenoids. So for the ECCS valves to be 18 operable they need to be closed and capable of opening 19 and that is the operability requirement straight from 20 the tech spec bases.
21 So, you know, technically I guess you 22 could say that one trip valve was not energized but 23 one was and in that case the ECCS valve would still be 24 operable.
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60 MEMBER ROBERTS: Okay. Thank you. Yeah, 1
in the meantime I was looking through the tech spec 2
document and that's consistent with the document that 3
operable is not really defined in that level of 4
detail, and so that does make sense.
5 MEMBER HALNON: Thank you, Tyler. Walt, 6
we are in your time.
7 MR. BOWMAN: I have one more thing to say.
8 MEMBER KIRCHNER: Dennis has his hand up.
9 Dennis, go ahead.
10 MEMBER HALNON: I don't think that's 11 Dennis. It's another one of the NuScale folks.
12 DR. BLEY: No, it's not me.
13 MEMBER HALNON: Doug from NuScale, do you 14 want --
15 MR. BOWMAN: This is Doug Bowman.
16 MEMBER HALNON: Go ahead, Doug.
17 MR. BOWMAN: Can you hear me?
18 MEMBER HALNON: Yes. Do you have a 19 clarification?
20 MR. BOWMAN: Yes. This is Doug Bowman.
21 I am the plant manager for Service, thank you, for 22 Services Operation, wow -- Services Manager for Plant 23 Operations at NuScale.
24 We did want to make one clarification for 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
61 the record. Monitoring during a control room 1
evacuation event occurs at the alternate operator work 2
stations and those are located at either the module 3
maintenance center or in the rad waste control room.
4 Thank you.
5 MEMBER HALNON: Thank you, Doug. Okay, 6
Walt, now it's to you.
7 MEMBER KIRCHNER: Matt, you are our lead 8
on this, have you any further questions of the Staff?
9 MEMBER SUNSERI: Thank you, Walt. I don't 10 have any. I think the Committee has asked all the 11 appropriate questions. Thanks.
12 MEMBER KIRCHNER: Other members?
13 (No response.)
14 MEMBER KIRCHNER: Well, then at this point 15 we have come to a logical break point in our schedule, 16 and so let's take a break until 10:15. That will 17 allow those of us on Mountain and Pacific Time to get 18 some coffee and refuel. And we'll reconvene at 10:15 19 and we'll take up the LOCA TR. Thank you.
20 (Whereupon, the above-entitled matter went 21 off the record at 9:59 a.m. and resumed at 10:14 a.m.)
22 CHAIR KIRCHNER: Okay. The meeting will 23 come back to order, and we are going to turn to the 24 topic of the Loss of Coolant Accident Topical Report 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
62 and turn to NuScale.
1 Sarah, are you ready?
2 MS. TURMERO: Yes. Good morning.
3 CHAIR KIRCHNER: Good morning. Go ahead.
4 MS. TURMERO: All right. Thank you. My 5
name is Sarah Turmero. I'm a licensing engineer for 6
NuScale covering topics on Chapter 4, 9, 15, and the 7
related topical reports. I've been with NuScale for 8
about two-and-a-half years and have a background in 9
PWR reactor engineering.
10 And with me, I have Meghan McCloskey and 11 Ben Bristol from the System Thermal Hydraulics Group 12 to assist with any questions if needed.
13 We'll be covering a summary of the 14 significant changes since the approval of the Revision 15 2 LOCA Topical Report. These changes include those 16 related to the scope, relevant design changes from 17 NPM-160 to NPM-120, and changes to the phenomena 18 identification and ranking table evaluation model 19 structure assessment basis updates and adequacy 20 assessment updates.
21 So the LOCA analysis method for a pipe 22 break inside containment and the associated event 23 classification figure of merits and key regulations 24 were maintained from the approved topical report. The 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
63 scope that was modified from the approved topical 1
report includes the incorporation of the analysis 2
method for the inadvertent opening of a reactor valve 3
scenario. And with that the event classification 4
figure of merit and associated key regulation was also 5
incorporated into the scope. Some key updates 6
associated with the IORV analysis is the 7
implementation of the new NSPN1 critical heat flux 8
correlation and modeling cross-flow between the hot 9
and average channel, and of course the scope that is 10 associated with the valve opening and inadvertent ECCS 11 actuation.
12 For NuScale from the approved topical 13 report the containment vessel pressure and temperature 14 response analysis methodology was incorporated into 15 this topical report. It was previously a separate 16 technical report and the associated figures of merit 17 and key regulations were also included as it relates 18 to the containment response methodology.
19 Additionally, the response to the LOCA pipe break, 20 secondary line breaks, and valve opening events --
21 those are added scope specifically related to 22 crediting DHRS.
23 Scope that is outside of the LOCA Topical 24 Report related to long-term cooling and subcriticality 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
64 are covered in the Extended Passive Cooling and 1
Reactivity Control Topical Report.
2 And the figure on the left shows how we've 3
defined phases of the event progression for the LOCA 4
and IORV events with Phase 0 being where NSPN1 is 5
implemented, and it's the first 10 seconds of the 6
transient. And then it occurs in conjunction with 7
Phase 1.
8 Since the DCA submittal, NuScale made 9
incremental improvements to the design and analysis 10 methods, resulting in margin improvement from overly 11 conservative methods or assumptions while maintaining 12 the same level of safety. So we have the power uprate 13 with no significant changes to the module, system, 14 structures, and components. The operating conditions 15 listed are nominal conditions and changes are a result 16 of the power uprate. And then for the containment 17 vessel the design pressure and temperature increased 18 and the upper material was changed.
19 Next slide?
20 MEMBER MARTIN: Just real quick, for the 21 open session we're oftentimes quiet because we get to 22 jump on you during closed, but given that we just 23 talked about Chapter 16, the tech specs, could you 24 briefly give an overview how you integrated the tech 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
65 spec into safety analysis? I know in the report 1
itself obviously you have. You may kind of highlight 2
maybe the major kind of the usual suspects as far as 3
initial conditions. I'm not sure there was a lot of 4
mention of how it relates back to tech specs, but of 5
course I know that they do. But I wanted to give you 6
the opportunity to just kind of talk about how you 7
incorporated the uncertainties which ultimately get 8
integrated into tech specs and into your initial 9
conditions, and that gets vetted to your safety 10 analysis.
11 MS. TURMERO: I was going to ask a 12 clarifying. Is there a specific tech spec or in 13 general?
14 MEMBER MARTIN: Well, maybe in closed 15 session.
16 MS. TURMERO: Okay.
17 MEMBER MARTIN: But just again this is a 18 public meeting. We just had a discussion on tech 19 specs and I thought it might be appropriate just to 20 segue from one to the other with this point, otherwise 21
-- I didn't want to save all my questions for closed 22 session.
23 MS. TURMERO: I think I'll start and then 24 we'll ask Karl to jump in from the tech spec side of 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
66 things.
1 But from the safety analysis perspective, 2
our initial conditions and the ranges of initial 3
conditions and the SSCs that we credit in the safety 4
analysis are largely consistent between the DCA and 5
the SDA, and we focused -- and the design changes have 6
changed operating conditions for design limits to 7
accommodate the increased power, but that structure is 8
largely the same. And then the one -- the new piece 9
of the system is related to the supplemental boron in 10 the ECCS that was incorporated into the tech specs.
11 Ben or Karl, do you want to add to that?
12 MR. BRISTOL: Sure, this is Ben Bristol.
13 So I think generally the flow goes the other direction 14 in our view, so safety analysis works pretty closely 15 with the design team on understanding the constraints 16 around the actual module itself as well as the system 17 team with the constraints around power production 18 targets things of that nature. And then we integrate 19 that with the I&C team, right, that helps set up and 20 establish what types of measurements we have, what 21 protections then we can derive from that. And all of 22 that package then goes into what ends up in tech 23 specs. Because it's a PWR, it largely looks a lot 24 like PWR tech specs.
25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
67 But we work closely with the tech spec 1
group on defining where those operational boundaries 2
are that accommodates the safety margins that we 3
demonstrate along with the consideration of the 4
required operational margins necessary to support the 5
plant systems and power production requirements.
6 MEMBER MARTIN: And there are a couple 7
other facets. Tech specs, some people might say, oh, 8
well, LOCA drives tech specs. That's not completely 9
true. LOCA is a limiting for all things, right? So 10 you get a mix of LOCA and non-LOCA informing tech 11 specs. When it comes to initializing for safety 12 analysis, you also -- contemporary approach is 13 following PIRT. You'll talk about that of course.
14 And not everything would justify biasing with all the 15 uncertainties and best estimates.
16 So part of the answer I guess I was kind 17 of probing was something to say that, well, you know, 18 we look at the PIRT and we kind of look at what's 19 important with regard to the influences coming in from 20 the initial conditions and we select -- you don't have 21 to bias everything because then it kind gets to become 22 a administrative nightmare if you try to be too cute 23 about it. But I was looking for a if-we-do-24 everything-kind of
- answer, but at least an 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
68 acknowledgement that there are priorities when it 1
comes to initializing the problem and it ties back to 2
obviously the physics of what to -- the problem that 3
you're solving and their influences throughout the 4
event. But that's fine.
5 MR. BRISTOL: Yes, and to build on that, 6
Sarah mentions the three different avenues of analysis 7
that's covered by the LOCA TR being the pipe break 8
scenarios and their figures of merit, which are 9
different than the containment analysis and its 10 figures of merit or the IORV in the short-term 11 transient core response figures of merit. And the 12 bias is -- the conservative bias directions are not 13 consistent between those three different (audio 14 interference).
15 MEMBER MARTIN: Of course. You have to 16 reconcile that sometimes, yes. Thanks.
17 MS. TURMERO: Next slide? All right. So 18 there were ECCS actuation signal modifications. So 19 with decreased RCS inventory, ECCS actuates on riser 20 level early in the transient progressions. The Tcold 21 interlock prevents ECCS actuation for extended DHRS 22 cooldown events. And the RCS level indication of 23 decreasing RCS inventory can generate ECCS signal 24 before significant containment level increase.
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69 To highlight some of the ECCS valve design 1
changes, the inadvertent actuation block was removed 2
from the vent valves and flow venturis were added to 3
the reactor vent and recirculation valves and the 4
third reactor vent valve was removed.
5 For long-term cooling, which is covered 6
with the scope of the Extended Passive Cooling Topical 7
Report, one of the relevant design changes is the 8
lower pool level. And even with the lower pool level, 9
the containment surface area below the pool level 10 provides ample core cooling and maintains ECCS 11 cooling.
12 DR. SCHULTZ: Sarah, this is Steve 13 Schultz. Could you just provide a general overview?
14 A lot of changes that you've described here, they just 15 came about as good ideas or did they come about as a 16 result of analysis evaluations that pertain to the 17 uprate?
18 MS. TURMERO: I can't speak to the 19 specific changes, but the improvements in the design 20 and analysis methods were done to gain margin 21 improvement, improve our analysis methods so that we 22 could spread margin across the design, operation, and 23 analysis.
24 But if you would like to -- Meghan?
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70 DR. SCHULTZ: Could you speak also to the 1
design changes themselves?
2 MEMBER MARTIN: I had a very similar 3
question, Steve.
4 PRA. Some changes are going to be 5
strictly to accommodate the power uprate and other 6
changes, given that of course you had a mature PRA at 7
DCA, that you might be able to go here I have an 8
opportunity to take advantage of this new insight.
9 And that could be another set of changes of a 10 different sort. So I think that was the perspective 11 that I was coming from was a very similar idea that 12 Steve had.
13 MR. BRISTOL: This is Ben Bristol again.
14 So we have a list. Maybe I'll just start with the 15 top. The removal of the IAB from the vent valves.
16 This was a key safety improvement that we found. Now 17 there are trade-offs. Depressurizing the vessel from 18 high pressure is something that we add -- that we 19 weigh very heavily. However, what we found -- Bob 20 mentioned the PRA insights -- the ability to actuate 21 ECCS on demand was a feature that we had precluded to 22 some degree with the IAB in the DCA design.
23 So it was a big improvement for accident 24 scenarios to have some assuredness on the timing of 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
71 ECCS actuation. There were event sequences where we 1
had a broad range of uncertainty based on when the IAB 2
would release due to its pressure-based lockout-3 blockout-type physics.
4 Generally for safety, right, we want lots 5
of water over the core, so we recognize that keeping 6
the IABs on the recirc valves was a good feature.
7 However, also part of safety is being able to 8
depressurize the reactor on demand. So removal of the 9
IABs from the vent valve allows us to depressurize the 10 reactor on demand, which allows us to reduce the 11 uncertainty of the timing of ECCS for a broad range of 12 events. And it really improves the response for 13 certain event sequences that we found to be very 14 beneficial. So that's just one. I could keep going.
15 Maybe another one of interest is the pool 16 level change that gets a lot of kind of consideration, 17 right? Seems like higher pool level would be better 18 for safety. As it turns out, the containment surface 19 area is ample. And we had plenty of heat removal 20 capability in the containment with margin to reduce 21 the pool level.
22 What we recognized in the design is that 23
-- you'll notice on the module here, we've got these 24
-- the big covers are access ports. And so one of the 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
72 challenges we had with the other design was the 1
ability to get in and service the equipment on top of 2
the reactor head. So dropping the pool level allowed 3
us to add those access ports that really improved the 4
maintainability of the design and some of the required 5
sequences.
6 And back to a different consideration of 7
safety, right, ALARA considerations where operators 8
are getting dose, getting them in and out of the 9
vessel effectively, efficiently was something that we 10 recognized is also a consideration of safety. And so 11 that was one of the trade-offs there, where we took 12 some of the margin that existed in the design from a 13 safety perspective and added it to a different element 14 of consideration of the design.
15 MEMBER HALNON: So, Ben, this is Greg.
16 How did you deal with the uncertainty at the tail end 17 of an accident when you're needing that extra volume 18 and (audio interference)?
19 MR. BRISTOL: Yes, that's a good question 20 and gets actually to one of the other bullets there is 21 the venting capacity. Very late in the design, we 22 depressurized the whole vessel down to sub-atmospheric 23 conditions. The containment effectively is a big 24 condenser, and so it can really draw the pressure 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
73 down. The physics of what happens there is it 1
actually adds stress to the venting capacity of the 2
vent valves. So the long-term core level response is 3
directly related to the flow or the pressure drop 4
across the vent valves.
5 So having a reduced pool level actually 6
allowed us some more margin in the sizing of the 7
venting needs, and that's what allowed the removal of 8
one of those valves, which has knock-on effects of 9
improving maintenance in space on top of the reactor 10 vessel with two valves instead of three and simplifies 11 the design, the number of components. It also reduces 12 the effect of the depressurization transient we talked 13 about if we were to have an inadvertent ECCS 14 actuation.
15 MEMBER MARTIN: Just a clarification. As 16 a matter of fact, you brought up that access ports 17 facilitate inspections and maybe maintenance on the 18 top. Now they have the RRBs below. Is that also 19 accessible for inspection at least and maybe some 20 maintenance if there was any issue down there?
21 MR. BRISTOL: Yes, so as part of refueling 22 we have a space where the upper module inclusive of 23 the recirc valves goes over to a dry dock essentially.
24 And I think most of those maintenance activities are 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
74 performed there.
1 MEMBER MARTIN: Okay.
2 MR. BRISTOL: But that's about as deep as 3
I can go on that topic.
4 MEMBER MARTIN: Okay. Yes, that's why --
5 you brought it up, so I -- you opened the door.
6 MR. BRISTOL: Sure. Most of the 7
components are located on the reactor head, which is 8
sort of where we identified that optimization.
9 DR. SCHULTZ: Thank you, Ben and Meghan.
10 That additional information is very helpful. Thank 11 you.
12 MS. TURMERO: Next slide, please? The 13 NuScale PIRT was reviewed for the NPM focusing on 14 topics like the break spectrum comparison, some 15 scaling analyses. The PIRT panel convened for a 16 focused evaluation on the phenomena associated with 17 valve opening events during initial rapid 18 depressurization. And regarding the impact of design 19 changes to LOCA, NuScale evaluated changes to the PIRT 20 geometric parameters and system state parameters, such 21 as pressures and temperatures, and found that these 22 changes did not introduce new phenomena or 23 significantly impact phenomena ranges.
24 DR. SCHULTZ: Sarah, this is Steve 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
75 Schultz. Was there overlap between the first PIRT 1
panel and this one, or was it the same panel that you 2
invited through the process?
3 MS. McCLOSKEY: There were a couple of 4
different PIRT panels as the design evolved and its 5
LOCA methodology evolved. I believe that the first 6
PIRT panel was actually in 2010. And then it was 7
updated in 2013 and updated in 2015. And so some of 8
the folks who were involved in the 2015 work were also 9
involved in the IORV-focused PIRT as well as the 10 review of the LOCA PIRT. So there was a little bit of 11 overlap. It wasn't a total reconvene of all of the 12 members. The updates were -- the PIRT panel members 13 were all internal to NuScale at this point in the 14 update space.
15 DR. SCHULTZ: So in this case, they 16 reviewed your -- NuScale's conclusions first that you 17 listed here, and then they looked specifically at the 18 topics that you provided?
19 MS. McCLOSKEY: I'd say we went the other 20 way around in terms of looking at the design changes 21 and the event progressions and the body of work that 22 NuScale had developed in terms of the NPM-160 work 23 that was done to support the DCA submittal and review.
24 And then we could build on that and compare it to the 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
76
-- what the break spectrums looked like with the 1
updated design and all the design changes incorporated 2
and evaluate whether there were significant changes 3
that warranted an update in the PIRT space.
4 DR. SCHULTZ: And in the last bullet, the 5
methodology changes identified, is that meaning that 6
the PIRT panel identified some changes that you then 7
implemented and evaluated, or is that something that 8
they evaluated?
9 MR. BRISTOL: So just to clarify, the IORV 10 phenomena were reviewed by a NuScale internal --
11 internally staffed PIRT panel.
12 DR. SCHULTZ: Okay.
13 MR. BRISTOL: The LOCA PIRT was reviewed 14 by our team essentially looking back through what the 15 PIRT panel had originally identified, the basis of 16 those rankings compared with the updated analysis 17 results to confirm that they were consistent. And the 18 limited scope methodology changes -- we'll get into 19 more detail on that in the closed session.
20 DR. SCHULTZ: Fine.
21 MR. BRISTOL: I don't think they were 22 primarily driven from phenomena rankings. Largely 23 they were driven by either needs of the power uprate 24 and assessment basis for areas where the margins had 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
77 required pencil sharpening or margin improvement.
1 DR. SCHULTZ: That helps. Thank you.
2 MS. TURMERO: Next slide, please? For the 3
DCA, NRELAP5 Version 1.4 was approved, and the current 4
evaluation model uses Version 1.7. The NIST-2 test 5
facility was upgraded -- the NIST-1 test was upgraded 6
to NIST-2, which allowed us to expand the NRELAP5 7
assessment bias by using the NIST-2 LOCA and IORV test 8
series.
Additional benchmark calculations or 9
sensitivity cases were performed as needed to support 10 these evaluation model changes.
11 MEMBER MARTIN: Question. Bob again. I 12 saw on their topical that the RELAP5 that you received 13 from Idaho is Version 4.13, one I'm familiar with.
14 It's also 13 years old, and there's been many updates 15 since that time. And I've worked with that code. I 16 know some of the limitations and that they were 17 resolved in some of the later condensations. Always 18 a challenge, particularly under low pressure. Of 19 course, you own NRELAP5. I mean, so as part of these 20 updates are you working with Idaho or are you --
21 you're taking their changes and incorporated it. So 22 it's really not 4.13 anymore. It actually embodies 23 some of the newer versions, and you're kind of keeping 24 up with what's going on with the development at Idaho.
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78 I see you nodding, but you can go ahead 1
and say it for the record.
2 MS. McCLOSKEY: Yes, so I'll say for the 3
record we -- I know that we do continue to get the bug 4
reports and code fixes that Idaho -- INL publishes to 5
the users group and incorporate that as part of our 6
normal (audio interference).
7 MEMBER MARTIN: All right. Some would 8
certainly be more important than the others if your 9
code was crashing. I mean, it's a reality of working 10 with system codes. I think it would be challenging.
11 And low pressure has always just been a huge 12 frustration. They're a lot better than they were when 13 we all started.
14 MS. TURMERO: Next slide, please? For the 15 evaluation model adequacy assessment, the bottom-up 16 and top-down evaluations that were performed for the 17 NPM-20 builds on the previously approved LOCA adequacy 18 assessment and the non-LOCA evaluation model 19 development for the steam generator and DHRS heat 20 transfer phenomena.
21 The top-down scaling analyses demonstrate 22 the important PI group similarity between the NPM-160 23 and NPM-120. There were no significant changes to the 24 field equations or numerical solutions and NRELAP5, 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
79 with the overall conclusion being that NRELAP5 and the 1
updated evaluation model are applicable and adequate 2
for our defined scope.
3 Next slide? To conclude, the updated LOCA 4
Topical Report describes the evaluation model using 5
NRELAP5 to analyze the NPM-20 LOCA and valve opening 6
events for Phase 0 and Phase 1A/B and the secondary 7
pipe break for the containment pressure and 8
temperature response.
9 We've covered a high-level summary of 10 relevant design changes that drive changes to NRELAP5, 11 including an expanded validation basis using the NIST-12 2 tests, and overall the LOCA Topical Report continues 13 to provide a robust methodology to analyze the NPM 14 response to LOCA valve opening events and the 15 containment pressure and temperature response 16 analysis.
17 With that, are there any additional 18 questions?
19 MEMBER HALNON: Walt, I don't see any 20 questions.
21 CHAIR KIRCHNER: Okay. Greg, then I think 22 at this point if there are no questions from the 23 members, we would turn to the staff for their open 24 presentation on the TR.
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80 MR. SNODDERLY: Walt, we're going to --
1 Tom Griffith is going to give us the HITI open slides, 2
and then we're going to go to the staff since 3
NuScale's all set up right now.
4 CHAIR KIRCHNER: Okay. Yes, that's more 5
efficient. Okay. Thank you, Mike.
6 MEMBER HALNON: Go ahead, Tom.
7 MR. GRIFFITH: Thank you. Thomas 8
Griffith, Licensing Manager of NuScale Power.
9 Just a little bit of background about 10 myself. Roughly 15 years' experience in the nuclear 11 industry. I held former positions as a senior reactor 12 operator, I&C manager, worked in safety analysis, 13 reactor engineering, and now work for -- in licensing 14 at NuScale. In charge of the US460 standard design 15 approval application.
16 What I intend to present here in the open 17 session is some high-level updates on the high-impact 18 tactical issues. There is a set of slides for the 19 closed session, where we can discuss some of the 20 aspects than what I'm going to present right now in 21 more detail.
22 Next slide, please? So if you recall, in 23 August when we discussed last the high-impact 24 technical issues. We have not identified any new 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
81 high-impact technical issues since that time. And at 1
this point there are effectively two high-impact 2
technical issues: No. 2 and 10 related to LOCA break 3
spectrum that we continue to work through defining a 4
clear path forward.
5 I did mark on this slide that the high-6 impact technical issue related to the IFR design 7
changes and the ASME qualification, the helical coil 8
steam generator, considered resolved by NuScale and 9
NRC management, that decision does officially take 10 place during our quarterly meetings which is next 11 week, but my understanding in discussions with my 12 counterparts at the NRC is that both sides do 13 recommend closure of the item, and enhanced with the 14 timing here in the presentation I wanted to at least 15 highlight that.
16 The high-impact technical issues related 17 to DWO we would consider resolved and we look forward 18 to presenting the material related to DWO here in some 19 of the upcoming ACRS meetings.
20 Next slide, please? So I do want to take 21 an opportunity here to provide an overview of the 22 approach to DWO. The purpose of this is kind of to 23 start some conversation. And we do have closed slide 24 presentations that accompany this presentation 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
82 material, but for the purpose of transparency in the 1
open session I want to walk through kind of what --
2 how NuScale views the resolution of DWO.
3 And so effectively what I'm showing here 4
is that NuScale's approach to DWO -- and we're calling 5
it our safety case -- by establishing three separate 6
pillars. We consider those pillars to be analysis, 7
real-time monitoring, and physical inspection. And 8
our approach is not limited to one specific area. So 9
for example, in analysis we've defined a DWO 10 transient. We have analyzed the steam generator 11 integrity with that transient, and we've defined a 12 time under which the steam generator can handle the 13 transient that we've analyzed.
14 We have also defined real-time monitoring, 15 which is -- I think I would characterize it as -- it 16 was a lot of feedback that we've gotten from both the 17 staff, our internal NuScale individuals, and in 18 discussion that we had in August with the ACRS. We 19 took a hard look at what we were providing to 20 operators, and we've now defined a pillar that we're 21 calling real-time monitoring. And that's effectively 22 a way for the operating team to infer where the steam 23 generator is operating with respect to DWO.
24 And then the last piece is physical 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
83 inspections. And so to accompany the analysis of the 1
real-time monitoring that's provided we've also 2
defined a set of inspections in intervals to provide 3
another layer of confidence that the steam generator, 4
being that it's a reactor coolant pressure boundary --
5 its integrity is maintained.
6 Next slide, please? So as I stated 7
before, we'll walk through some of the pillars a 8
little bit here. I do have more detailed slides in 9
the closed session, but effectively under analysis, 10 like I said, we've defined a DWO transient. We've 11 done the evaluations with the transient to demonstrate 12 steam generator structural integrity. We've defined 13 real-time monitoring using a comparison between our 14 RCS hot temperature and main steam temperatures, which 15 are safety-related indications. And then we've also 16 defined a limit, and that limit is required by tech 17 specs for how long a particular steam generator could 18 operate in a region where there's the potential for 19 DWO.
20 Next slide?
21 MEMBER MARTIN:
Bob Martin.
A 22 clarification. My understanding is that with the DWO, 23 you were going to more or less design it out of 24 normal/abnormal operation or design-basis conditions.
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84 And of course part of that would be you would have 1
protection systems to otherwise ensure that this 2
doesn't happen, as opposed to conditions, or allowing 3
for conditions that would result in DWO, right? To 4
clarify, basically take it out of a DBA space or DBE 5
space. Is that correct? Do I understand that 6
properly?
7 MR. GRIFFITH: So a current approach to 8
DWO, the way I would define it, is that we have 9
established what we -- a representative transient for 10 DWO and analyzed the steam generator for a particular 11 time frame that that transient could occur.
12 MEMBER MARTIN:
Could occur with 13 assumptions that make it a DBE-kind of thing, or is it 14 a beyond-design-basis condition?
15 MR. GRIFFITH: I think some of the 16 specifics you -- we would want to get into are 17 probably more appropriate for closed, but I would say 18 that the loads induced from the phenomena are 19 relatively low and not impactful. In fact, when we 20 look at the expected operation -- and we have the 21 figure up right now that has been placed in the FSAR 22 5.1-16 -- we expect the majority of operation to occur 23 in Region II, and Region II is defined as the region 24 that precludes DWO.
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85 That does not mean -- and I don't want to 1
misrepresent -- Region I does not mean you have DWO.
2 Region I simply is that there is less margin to DWO.
3 And there are some slides in the closed session that 4
walk through exactly how those margins are defined.
5 But what we've done is in that Region II -- even if --
6 based upon the total amount of time that we would 7
expect an applicant to operate, there is sufficient 8
margin that roughly seven years or so of additional 9
margin that DWO could continue or could occur 10 continuously before hitting our acceptance criteria 11 for wear.
12 So what I'm trying to say is that 90-13 percent-plus of the operation would be in this DWO is 14 precluded. We had to set an analysis limit somewhere 15 with acceptance criteria. The amount of margin in 16 there before we would -- we would say internally, hey, 17 we need to look at this further is such that there --
18 or there's on the order of multiple years or many 19 years. And it's kind of tricky because there's 20 different impacts from DWO like, you know, for 21 example, sliding wear is different than thermal 22 fatigue. So I think that we need -- we would need to 23 get into some of those finite details, but there is 24 substantial margin before DWO would cause any sort of 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
86 concern with the steam generator integrity.
1 MEMBER MARTIN: I guess where I was going 2
with it, although your answer is what I expected, 3
we've gone through a lot of discussion on design 4
changes. Were there any design changes to support the 5
strategy and specifically related to I&C monitoring 6
that might result in some sort of protective action?
7 MR. GRIFFITH: I think I can get into that 8
in the closed session.
9 MEMBER MARTIN: Okay. I mean yes or no 10 might work, but --
11 MR. GRIFFITH: To some extent the answer 12 is yes.
13 MEMBER MARTIN: Good enough.
14 MR. GRIFFITH: Next slide, please? So 15 lastly, the other piece we'd talked through a little 16 bit is we've set a number of physical examinations 17 that include the steam generator and associated 18 components and specified frequencies that would inform 19 what future inspections may need to look at and 20 provide the assurance of steam generator integrity.
21 Next slide?
22 MEMBER HARRINGTON: One question real 23 quick. This is Craig Harrington. Are you envisioning 24 that inspection schedule as specific to the first 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
87 module that operates or something that would be 1
appropriate long term?
2 MR. GRIFFITH: So we do have specifics on 3
the first module.
4 Erin, I don't know if you can step here on 5
the exact language for the first module?
6 But there are specifics for the first 7
module under operation. And the frequency is set to 8
complement what we've done for analysis.
9 MS. BLUMSACK: Yes, this is Erin Blumsack 10 from NuScale. For the first NPM that undergoes a 11 refueling outage, after the first 100 percent tube 12 examinations at the first refueling outage they're 13 required to inspect at least 20 percent of the steam 14 generator tubes at each outage with the tech spec 15 requirement that they have to get to 100 percent of 16 tube inspection by 72 EFPM after the first outage.
17 That is a COL item in Chapter 5. Tech specs have not 18 changed and that's only for the first NPM that 19 undergoes refueling.
20 MEMBER BALLINGER: This is Ron Ballinger.
21 I'm not sure -- I don't remember, so I'll ask the 22 question: Since it's externally pressurized tubes, a 23 lot of the inspections for various phenomena that 24 occur in a commercial PWR -- it's not applicable. But 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
88 there are others. For example, dimensions. Is there 1
inspection going to occur that looks at the ID? In 2
other words, just to find out if there's any kind of 3
pre-collapse or creep-down that's going on in the 4
tubes?
5 MS. BLUMSACK: Details of examinations 6
will be developed for a COL applicant as part of the 7
Steam Generator Program.
8 MEMBER BALLINGER: I guess -- okay.
9 MS. BLUMSACK: Does that address your 10 question?
11 MEMBER BALLINGER: Yes, I guess. We have 12
-- a lot of us have issues every time somebody says 13 that's up to the COL applicant, so --
14 MS. BLUMSACK: Understood. We expect to 15 be able to use examination techniques that the 16 industry uses, but that will be developed in more 17 detail during the Steam Generator Program.
18 MEMBER BALLINGER: You'll find out the 19 first time a bobbin coil gets stuck in one of the 20 tubes.
21 MS. BLUMSACK: That is true.
22 MR. GRIFFITH: And I believe that that was 23 the last slide that I had, so if there's any further 24 questions for the open session, any update?
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89 MEMBER HALNON: Walt, I don't see anything 1
in the room.
2 CHAIR KIRCHNER: Other members, any 3
questions?
4 (No response.)
5 CHAIR KIRCHNER: Okay. Bear with me. I'm 6
on a small NRC computer screen. I have to move so I 7
can look at the agenda, so bear with me.
8 Okay. With that then, I believe we are 9
ready, Mike, to go to the staff's evaluation of the 10 LOCA TR in the open session.
11 MR. SNODDERLY: Yes, sir, that's --
12 CHAIR KIRCHNER: Is that correct?
13 MR. SNODDERLY: I just need a minute to 14 switch presenters.
15 MEMBER HALNON: I'll let you know when 16 we're ready, Walt.
17 CHAIR KIRCHNER: Okay. So when you're 18 ready, just go ahead. I can't see --
19 MR. SNODDERLY: I'll let you know. And 20 then also an opportunity for public comment after the 21 staff's presentation. And then --
22 CHAIR KIRCHNER: Yes.
23 MR. SNODDERLY: -- that will be end of the 24 open session.
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90 CHAIR KIRCHNER: Right. Correct. Okay.
1 MEMBER HALNON: Okay. Just give us a 2
minute.
3 (Pause.)
4 CHAIR KIRCHNER: Okay. Staff, whenever 5
you're ready.
6 MR. VIVANCO: Good morning, everyone. My 7
name is Ricky Vivanco, and I'm a project manager in 8
the NRR New Reactor Licensing Branch. I am the PM 9
assigned to the staff's review of NuScale's Loss of 10 Coolant Accident Evaluation Model Topical Report.
11 The technical reviewers a part of this 12 review are: Dr. Shanlai Lu, Dr. Sean Piela, Dr. Dong 13 Zheng, Mr. Carl Thurston, Mr. Ryan Nolan, Dr. Syed 14 Haider, Dr. Joshua Kaizer, Dr. Peter Lein from the 15 Office of Research, and Dr. Leonard Ward from Numark.
16 Again, I am the project manager assigned to this 17 topical report supported by (audio interference) for 18 the overall project.
19 A review of this topical report. The 20 Revision 3 of the LOCA Evaluation Topical Report was 21 submitted on January 5th, 2023, and the topical report 22 was accepted for review on July 31st, 2023.
23 The staff conducted an audit from March 24 2023 to August 31st, 2024. Fifty-seven audit issues 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
91 were generated resulting in supplemental information 1
submitted by NuScale. Of those items not resolved 2
during the audit, four RAIs were generated. I want to 3
be clear here, all audit items and RAIs were 4
resolved/closed that were specifically related to the 5
LOCA Topical Report. I do want to point out though 6
there are confirmatory items that are awaiting 7
confirmation and upcoming revisions or related to open 8
items in other areas of the review -- in other topical 9
reports.
10 Due to the technical and proprietary 11 nature of the topical report, the details of the 12 staff's review are going to be covered in the closed 13 session, however we will go over the conclusions here.
14 Subject to the closure of those open and 15 confirmatory items I noted, and along with 11 16 limitations and conditions identified, the staff 17 concludes that the methodology is acceptable for 18 meeting the requirements of 10 CFR 50.46 and the 19 associated portions of Appendix K evaluated in the 20 topical report.
21 For evaluation of the ECCS performance in 22 the NuScale NPM-20 for design-basis LOCAs, the 23 proposed LOCA evaluation model is conservative to 24 determine CHF and collapsed liquid level above the 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
92 reactor core. Further, the staff finds that the 1
containment response analysis methodology is 2
conservative and acceptable and that the NRELAP5 3
computer code and the NPM-20 model are acceptable to 4
evaluate the MCHFR and IORV in LOCA events.
5 And that's the end of the staff's 6
presentation. I'll defer any questions to Dr. Shanlai 7
Lu.
8 DR. LU: As our project manager mentioned, 9
there are a lot of details we can present in the 10 proprietary session, but if there are any questions 11 for the staff at this point in open session, I'm here 12 to answer.
13 MEMBER HALNON: I don't see any in the 14 room, Walt.
15 CHAIR KIRCHNER: Members online, any 16 questions?
17 Hearing none, I think we are at a juncture 18 where we should ask for any public comment.
19 So members of the public either in the 20 room or online -- just those of you online, un-mute 21 yourself and state your name and affiliation as 22 appropriate and ask your question. Or make your 23 comment. Excuse me. Not question.
24 MEMBER HALNON: We have no one in the 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
93 room, Walt.
1 CHAIR KIRCHNER: Okay. Online, any 2
comments from the public?
3 (No response.)
4 CHAIR KIRCHNER: Okay. I think, Greg, we 5
don't have any public input today.
6 With that, then, I think we're at the 7
juncture where we can close this open session and move 8
to an actual closed session. And that will take up 9
the LOCA Evaluation Model first.
10 So with that, for those of you that 11 attended on the open session, thank you.
12 Again for the record, I want to thank 13 NuScale for joining us so early this morning. And 14 this open session is closed. Thank you.
15 MEMBER HALNON: Okay. We'll be logging 16 off this one and be logging on -- there should be a 17 new link for those that are invited to the closed 18 session. This session will be closed.
19 (Whereupon, the above-entitled matter went 20 off the record at 11:04 a.m.)
21 22 23 24 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1716 14th STREET, N.W., SUITE 200 (202) 234-4433 WASHINGTON, D.C. 20009-4309 www.nealrgross.com
LO-177832 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com January 09, 2025 Docket No. 052-050 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738
SUBJECT:
NuScale Power, LLC Submittal of Presentation Material Entitled ACRS Subcommittee Meeting (Open Session) Chapter 16, Part 4, LOCA LTR, and HITI Status, PM-177830, Revision 0 The purpose of this submittal is to provide presentation materials for use during the upcoming Advisory Committee on Reactor Safeguards (ACRS) NuScale Subcommittee Meeting on January 15, 2025. The materials support NuScales presentation of the subject chapter, topical report and status of the US460 Standard Design Approval Application.
The enclosure to this letter is the nonproprietary presentation entitled ACRS Subcommittee Meeting (Open Session) Chapters 16, Part 4, LOCA LTR, and HITI Status, PM-177830, Revision 0.
This letter makes no regulatory commitments and no revisions to any existing regulatory commitments.
If you have any questions, please contact Jim Osborn at 541-360-0693 or at josborn@nuscalepower.com.
Sincerely, Thomas Griffith Director, Regulatory Affairs NuScale Power, LLC Distribution:
Mahmoud Jardaneh, Chief New Reactor Licensing Branch, NRC Getachew Tesfaye, Senior Project Engineer, NRC Michael Snodderly, Senior Staff Engineer, Advisory Committee on Reactor Safeguards, NRC : ACRS Subcommittee Meeting (Open Session) Chapters 16, Part 4, LOCA LTR, and HITI Status, PM-177830, Revision 0
LO-177832 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com ACRS Subcommittee Meeting (Open Session) Chapters 16, Part 4, LOCA LTR, and HITI Status, PM-177830, Revision 0
1 PM-177830 Rev. 0 Copyright © 2025 NuScale Power, LLC.
NuScale Nonproprietary Template #: 0000-21727-F01 R10 ACRS Subcommittee Meeting (Open Session)
January 15, 2025 Chapter 16, Part 4, LOCA LTR and HITI Status
2 PM-177830 Rev. 0 Copyright © 2025 NuScale Power, LLC.
NuScale Nonproprietary Template #: 0000-21727-F01 R10 ACRS Subcommittee Meeting (Open Session)
January 15, 2025 Presenter Gene Eckholt Chapter 16 Technical Specifications
3 PM-177830 Rev. 0 Copyright © 2025 NuScale Power, LLC.
NuScale Nonproprietary Template #: 0000-21727-F01 R10 Acknowledgement and Disclaimer This material is based upon work supported by the Department of Energy under Award Number DE-NE0008928.
This presentation was prepared as an account of work sponsored by an agency of the United States (U.S.)
Government. Neither the U.S. Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the U.S. Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the U.S. Government or any agency thereof.
4 PM-177830 Rev. 0 Copyright © 2025 NuScale Power, LLC.
NuScale Nonproprietary Template #: 0000-21727-F01 R10 NuScale SDAA Part 4, US460 Generic Technical Specifications (GTS)
Subpart E of 10 CFR 52, Standard Design Approvals, does not require submittal of Technical Specifications for consideration.
In the Statements of Consideration for the 2007 rule change to 10 CFR Part 52, the commission expressed its expectation that the contents of applications for design approvals should contain essentially the same technical information that is required of design certification applications.
NuScale included Part 4, Generic Technical Specifications, in the SDAA.
5 PM-177830 Rev. 0 Copyright © 2025 NuScale Power, LLC.
NuScale Nonproprietary Template #: 0000-21727-F01 R10 Noteworthy US460 Design Changes Affecting GTS Rated thermal power increase from 160 MWt to 250 MWt Modification of ECCS design from three to two reactor vent valves Addition of ECCS supplemental boron system Addition of passive autocatalytic recombiner in containment
6 PM-177830 Rev. 0 Copyright © 2025 NuScale Power, LLC.
NuScale Nonproprietary Template #: 0000-21727-F01 R10 US460 GTS Development Started with NuScale US600 Certified Design Technical Specifications as model Addressed US460 design changes Applied 10 CFR 50.36 criteria to plant design, operations, and safety analyses Used industry STS Writers Guide format and guidance Incorporated recent industry STS changes as appropriate Technical Report TR-101310 Rev 0 describes the differences between US600 and US460 GTS at the time of SDAA submittal
7 PM-177830 Rev. 0 Copyright © 2025 NuScale Power, LLC.
NuScale Nonproprietary Template #: 0000-21727-F01 R10 Noteworthy GTS Changes Described in TR-101310 Revision 0 The MODE definition revised to better align with the plant response behavior The reactor core critical heat flux correlations and limits, and the RCS pressure safety limits revised to reflect the increased reactor power and changes to the plant design New Surveillance Requirement to ensure isolation of Module Heatup System between modules Module Protection System requirements modified to align with design changes Remote Shutdown Station LCO removed RCS Operational Leakage LCO and definition modified to align with industry standards to the extent appropriate for the NuScale Design
8 PM-177830 Rev. 0 Copyright © 2025 NuScale Power, LLC.
NuScale Nonproprietary Template #: 0000-21727-F01 R10 Noteworthy GTS Changes Described in TR-101310 Revision 0 (continued)
LTOP and ECCS LCOs modified to reflect reduced number of reactor vent valves UHS LCO modified to reflect design changes New LCO to ensure OPERABILITY of ECCS Supplemental Boron System New LCO to ensure containment closure during module movement between operating location and containment closure tool LCO 3.7.3 removed due to change from leak-before-break to break exclusion Chapter 5 Administrative Controls modified to reflect approved control room staffing plan
9 PM-177830 Rev. 0 Copyright © 2025 NuScale Power, LLC.
NuScale Nonproprietary Template #: 0000-21727-F01 R10 US460 GTS Review Audit Results 68 audit items resolved Most changes were editorial or clarifications Noteworthy changes included o
Core reactivity balance surveillance frequency clarified o
Module Heatup System flow paths added to Boron Dilution Control specification o
ECCS Supplemental Boron specification revised to include requirements for the geometric form of boron pellets RAI Results No RAI questions on Chapter 16 or GTS GTS change resulting from RAI associated with another FSAR Chapter o
Steam Generator Program revised to update the SG tube integrity discussion
10 PM-177830 Rev. 0 Copyright © 2025 NuScale Power, LLC.
NuScale Nonproprietary Template #: 0000-21727-F01 R10 Noteworthy GTS Change Not Associated with SDAA Review Added LCO to ensure OPERABILITY of passive autocatalytic recombiner (PAR) o NuScale determined the PAR mitigates design-basis events, making the component safety-related and appropriate for inclusion in the GTS
11 PM-177830 Rev. 0 Copyright © 2025 NuScale Power, LLC.
NuScale Nonproprietary Template #: 0000-21727-F01 R10 Acronyms CFR Code of Federal Regulations ECCS Emergency Core Cooling System FSAR Final Safety Analysis Report HVAC Heating, Ventilation, and Air Conditioning LCO Limiting Condition for Operation LTOP Low Temperature Overpressure Protection MWt Megawatts Thermal PAR Passive Autocatalytic Recombiner RAI Request for Additional Information RCS Reactor Coolant System SDAA Standard Design Approval Application STS Standard Technical Specifications TS Technical Specifications UHS Ultimate Heat Sink
12 PM-177830 Rev. 0 Copyright © 2025 NuScale Power, LLC.
NuScale Nonproprietary Template #: 0000-21727-F01 R10 ACRS Subcommittee Meeting (Open Session Session)
January 15, 2025 Presenter: Sarah Turmero Loss-of-Coolant Accident Topical Report
13 PM-177830 Rev. 0 Copyright © 2025 NuScale Power, LLC.
NuScale Nonproprietary Template #: 0000-21727-F01 R10 Agenda Summary of significant changes since TR-0516-49422-P Revision 2 approval o Analysis scope addressed by topical report o Design changes from 160 MWt NPM-160 design to 250 MWt NPM-20 design o Summarize effects on PIRT o EM structure and assessment basis updates o Adequacy assessment process and conclusions Conclusions
14 PM-177830 Rev. 0 Copyright © 2025 NuScale Power, LLC.
NuScale Nonproprietary Template #: 0000-21727-F01 R10 LOCA Topical Report: Analysis Purpose and Transient Class Scope maintained from DCA Topical Report:
o Loss-of-coolant accident (LOCA) pipe break inside containment analysis method o Event classification: Postulated accident o Figures of merit: Phase 1a, 1b collapsed liquid level (CLL) over top of fuel, minimum critical heat flux ratio (MCHFR) o Key Regulations: 10 CFR 50.46, 10 CFR 50 Appendix K, GDC 35 Scope modified from DCA Topical Report:
o Inadvertent opening of a reactor valve (IORV) analysis method o Event classification: conservatively classified as Anticipated Operational Occurrence (AOO)
Realistically not expected to occur during a module lifetime o Figure of merit: MCHFR during Phase 0 initial blowdown o Key Regulations: GDC 10 Scope added from DCA Topical Report:
o Containment vessel (CNV) pressure/temperature response analysis method
Similar to method used in DCA technical report o Response to LOCA pipe break, secondary line breaks, IORV events, or inadvertent ECCS actuation o Figures of merit: Maximum CNV pressure, maximum CNV wall temperature, CNV pressure reduction over time o Key Regulations: GDC 16, PDC 38, GDC 50 Scope addressed elsewhere:
o Extended passive cooling and reactivity control (XPC) topical report addresses long-term core cooling and subcriticality EM Capability Requirements
15 PM-177830 Rev. 0 Copyright © 2025 NuScale Power, LLC.
NuScale Nonproprietary Template #: 0000-21727-F01 R10 Power Uprate and Design Changes Summary NPM-160 to NPM-20 Power uprate from 160 MWt to 250 MWt Module SSC design essentially maintained Operating conditions o Increased primary pressure from 1850 psia to 2000 psia o Primary and secondary side design pressures increased from 2100 psia to 2200 psia o Use Tavg control instead of Thot control (RCS Tavg change from ~545°F to 540°F) o Decreased secondary side feedwater temperature at 100% power from 300°F to 250°F o Reduced minimum temperature for criticality from 420°F to 345°F Containment vessel o Design pressure increased from 1050 psia to 1200 psia o Design temperature increased from 550°F to 600°F o Upper containment material change from SA-508 to SA-336 F6NM EM Capability Requirements
16 PM-177830 Rev. 0 Copyright © 2025 NuScale Power, LLC.
NuScale Nonproprietary Template #: 0000-21727-F01 R10 Power Uprate and Design Changes Summary NPM-160 to NPM-20, contd ECCS actuation signals modified o Low RCS level (top of riser), Tcold interlock o Low-low RCS level (mid-riser) always active (no interlocks) o Timers:
8-hour timer on all reactor trips; operators can block the actuation if subcriticality at cold conditions is confirmed and combustible gas mixture in RPV is precluded
24-hour timer after loss of AC power supply (unchanged from DCA)
ECCS valve design changes o Removed IAB from vent valves to enhance depressurization capability in DBE and BDBE o Modified IAB threshold/release pressures on recirculation valves o Added second trip valve to each ECCS valve to prevent inadvertent opening on solenoid failure o Added flow venturi to RVVs and reactor recirculation valves (RRVs) o Removed third RVV Long-term passive cooling enhancements for collapsed liquid level and subcriticality FOM o Addressed by Extended Passive Cooling and Reactivity Control methodology o Design changes reduce but maintain ample CNV cooling capacity:
Lowered reactor pool level from ~68 ft to ~53 ft
Reduced conductivity in upper CNV due to material change o Mitigation of boron redistribution during DHRS and ECCS cooling with riser hole flow paths o Supplemental ECCS Boron (ESB) to maintain subcriticality during extended passive cooling EM Capability Requirements
17 PM-177830 Rev. 0 Copyright © 2025 NuScale Power, LLC.
NuScale Nonproprietary Template #: 0000-21727-F01 R10 Phenomena Identification and Ranking NuScale reviewed the NPM-160 LOCA PIRT for applicability to the 250 MWt NPM-20 design based on:
o Plant break spectrum comparisons o Scaling analyses o Other NPM-160 work subsequent to the PIRT development NuScale convened a PIRT panel for focused evaluation of IORV phenomena during rapid initial depressurization Overall conclusions:
o Existing phenomena identification and rankings applicable for 250 MWt NPM-20 design o Clarified phenomena importance based on work for NPM-160 and NPM-20 designs performed after development of NPM-160 LOCA PIRT Limited scope of methodology changes identified, and supported by additional NRELAP5 validation and sensitivity analyses EM Capability Requirements
18 PM-177830 Rev. 0 Copyright © 2025 NuScale Power, LLC.
NuScale Nonproprietary Template #: 0000-21727-F01 R10 EM Structure and Assessment Basis NRELAP5 v1.4 previously approved in DCA Current EM employs NRELAP5 v1.7 Upgraded the NIST-1 integral effects test facility to NIST-2 Expanded NRELAP5 assessment basis:
o NIST-2 LOCA test series o NIST-2 IORV test series Additional benchmark calculations, sensitivity cases as needed to support EM changes
19 PM-177830 Rev. 0 Copyright © 2025 NuScale Power, LLC.
NuScale Nonproprietary Template #: 0000-21727-F01 R10 EM Adequacy Assessment Bottom-up and top-down evaluations performed o Builds on LOCA EM adequacy assessment performed for DCA o Builds on the non-LOCA EM development for SG/DHRS heat transfer phenomena Compared NPM-160 and NPM-20 geometry, operating conditions, range of conditions for LOCA spectrum Evaluated scope of NRELAP5 code changes since v1.4 Top-down scaling analyses demonstrated important PI group similarity between NPM-160, NPM-20, NIST-2 No significant changes to NRELAP5 field equations or numerical solution NIST-2 LOCA and IORV tests expand the NRELAP5 assessment basis for NPM integral response
==
Conclusion:==
NRELAP and updated EM are applicable and adequate for the defined scope.
EM Adequacy Assessment
20 PM-177830 Rev. 0 Copyright © 2025 NuScale Power, LLC.
NuScale Nonproprietary Template #: 0000-21727-F01 R10 Conclusions Updated topical report describes evaluation model for use of NRELAP5 to analyze:
o NPM-20 LOCA or valve opening events, to assess Phase 0 MCHFR, Phase 1a/1b MCHFR, collapsed liquid level, o NPM-20 LOCA, valve opening, and secondary pipe break containment pressure response Design changes from NPM-160 to NPM-20 evaluated for effect on LOCA or valve opening transient and important phenomena NRELAP5 code changes incorporated as necessary to support the NPM-20 EMs NRELAP5 validation basis expanded with NIST-2 tests Topical report provides robust methodology to analyze NPM response to LOCA and valve opening events, and for containment pressure/temperature response analysis.
21 PM-177830 Rev. 0 Copyright © 2025 NuScale Power, LLC.
NuScale Nonproprietary Template #: 0000-21727-F01 R10 Questions?
22 PM-177830 Rev. 0 Copyright © 2025 NuScale Power, LLC.
NuScale Nonproprietary Template #: 0000-21727-F01 R10 Acronyms AC Alternating Current AOO Anticipated Operational Occurrence BDBE Beyond Design Basis Event CLL Collapsed Liquid Level CNV Containment Vessel DBE Design Basis Event DCA Design Certification Application DHRS Decay Heat Removal System ECCS Emergency Core Cooling System EM Evaluation Model ESB ECCS Supplemental Boron FOM Figure of Merit IAB Inadvertent Actuation Block IORV Inadvertent Opening of an RPV Valve GDC General Design Criteria LOCA Loss-of-Coolant Accident MCHFR Minimum Critical Heat Flux Ratio NIST NuScale Integral NPM NuScale Power Module PDC Principal Design Criteria PIRT Phenomena Identification and Ranking Table RAI Request for Additional Information SDAA Standard Design Approval Application SDA Standard Design Approval SSC Systems, Structures, and Components RCS Reactor Coolant System RPV Reactor Pressure Vessel RRV Reactor Recirculation Valve RVV Reactor Vent Valve XPC Extended Passive Cooling
23 PM-177830 Rev. 0 Copyright © 2025 NuScale Power, LLC.
NuScale Nonproprietary Template #: 0000-21727-F01 R10 ACRS Subcommittee Meeting (Open Session)
January 15, 2025 Presenter: Thomas Griffith Update - High Impact Technical Issues
24 PM-177830 Rev. 0 Copyright © 2025 NuScale Power, LLC.
NuScale Nonproprietary Template #: 0000-21727-F01 R10 High Impact Technical Issues (HITIs)
- 1. Design and classification of the augmented DC power system (EDAS)
- 2. Loss-of-Coolant (LOCA) break spectrum
- 3. Incorporated by reference (IBR)
- 4. Containment Vessel (CNV) material change
- 5. Lower reactor pressure vessel (RPV) material change
- 6. Secondary side controller design for density wave oscillation (DWO) events
- 7. DWO and steam generator inlet flow restrictor (IFR) design changes
- 8. ASME qualification of the helical coil steam generator for the onset of DWO-induced loads
- 9. Upper-to-lower RPV flange bolted joint shear loading that results from differential thermal expansion 10.LOCA break at CVCS/CIV connection Note: Green indicates issues that have been considered resolved by NuScale and NRC Management
25 PM-177830 Rev. 0 Copyright © 2025 NuScale Power, LLC.
NuScale Nonproprietary Template #: 0000-21727-F01 R10 Density Wave Oscillation Safety Case
- Analyses - DWO transient, which is used to assess SG structural integrity. SG structural integrity ensured for longer than NPM lifetime limit for time in DWO.
- Real-Time Monitoring - Defined operational space where DWO is precluded and where time in DWO is conservatively accounted for against the NPM lifetime limit.
- Physical Inspections - Examinations of SG tubes and IFRs ensure RCPB integrity is maintained. Degradation assessment will ensure that any damage to the tubes will inform future examination locations and frequencies.
26 PM-177830 Rev. 0 Copyright © 2025 NuScale Power, LLC.
NuScale Nonproprietary Template #: 0000-21727-F01 R10 DWO Safety Case (Continued)
- Analyses o DWO transient defined in SDAA FSAR Section 3.9.1 and lifetime limit specified in Table 3.9-1 o SG structural integrity is evaluated beyond the DWO lifetime limit for the NPM 60-year design life.
- Real-Time Monitoring o SG approach temperature Comparison between RCS hot temperature and main steam temperature o Time is counted in DWO against FSAR Table 3.9-1 Summary of Design Transients 60-year US460 design life limit of 2840 days in DWO.
Technical Specifications 5.5.3 cyclic limits
27 PM-177830 Rev. 0 Copyright © 2025 NuScale Power, LLC.
NuScale Nonproprietary Template #: 0000-21727-F01 R10 DWO Safety Case (Continued)
- Real-Time Monitoring o DWO is precluded during normal operations by maintaining an adequate SG approach temperature.
DWO is precluded in Region 2 Margin between normal operation and the Region 1/Region 2 boundary; Margin between the Region 1/Region 2 boundary and DWO onset.
Operation with DWO is avoidable for most of the NPM operating life.
28 PM-177830 Rev. 0 Copyright © 2025 NuScale Power, LLC.
NuScale Nonproprietary Template #: 0000-21727-F01 R10 DWO Safety Case (Continued)
- Physical Examinations o SG tube examination requirements in Technical Specifications 5.5.4 100 percent SG tube examination at first refueling outage 100 percent SG tube examination over 72 EFPM (~ 6 years) after first refueling outage:
- Maximum time below approach temperature boundary (2840 days or >7 years) is greater than maximum time between SG tube examinations.
- Additional requirement to inspect at least 20 percent of tubes per outage for the first NPM to undergo a refueling outage Degradation assessment program will ensure that examination results factor into future examination frequency and location.
VT-3 examination of IFRs
29 PM-177830 Rev. 0 Copyright © 2025 NuScale Power, LLC.
NuScale Nonproprietary Template #: 0000-21727-F01 R10 Acronyms ASME American Society of Mechanical Engineers CIV Containment Isolation Valve CNV Containment Vessel CVCS Chemical and Volume Control System DWO Density Wave Oscillation EDAS Augmented DC Power System EFPM Effective Full Power Months FSAR Final Safety Analysis Report HITI High Impact Technical Issue IBR Incorporate by Reference IFR Inlet Flow Restrictor LOCA Loss-of-Coolant Accident NPM NuScale Power Module RCPB Reactor Coolant Pressure Boundary RCS Reactor Coolant System RPV Reactor Pressure Vessel SDAA Standard Design Approval Application SG Steam Generator
Non-Proprietary Presentation to the ACRS Subcommittee Staff Review of NuScale SDAA Part 2, Chapter 16, Technical Specifications, and Part 4, US460 Generic Technical Specifications, Volume 1, Specifications, and Volume 2, Bases Revision 1 January 15th, 2025 (Open Session) 1
Non-Proprietary NuScale SDAA Part 2 Chapter 16 and Part 4 Review NuScale submitted Part 2 (FSAR), Chapter 16, Technical Specifications (TS), and Part 4, US460 Generic Technical Specifications (GTS), Revision 0, of the NuScale SDAA on December 29 and December 31, 2022, respectively, and Revision 1 on October 31, 2023 NRC regulatory audit of FSAR Chapter 16 and Part 4 was performed from March 2023 to August 2024, generating 68 audit issues All audit issues were resolved in the audit 52 audit issues resulted in NuScale submitting supplemental information to address questions raised during the audit No RAIs issued Staff completed review of FSAR Chapter 16 and Part 4 and issued an advanced safety evaluation to support today's ACRS Subcommittee meeting 2
Overview
Non-Proprietary NuScale SDAA Part 2 Chapter 16 and Part 4 Review Technical Reviewers
- Craig Harbuck, Lead Reviewer, NRR/DSS/STSB
- Steve Smith, NRR/DSS/STSB
- Clint Ashley, NRR/DSS/STSB
- Josh Wilson, NRR/DSS/STSB Project Managers
- Alina Schiller, PM, NRR/DNRL/NRLB
- Getachew Tesfaye, Lead PM, NRR/DNRL/NRLB 3
Contributors
Non-Proprietary NuScale SDAA Part 2 Chapter 16 and Part 4 Review Chapter 1.0 Use and Application Chapter 2.0 Safety Limits (SLs)
Chapter 3.0 Limiting Conditions for Operation (LCOs) and Surveillance Requirements (SRs)
Chapter 4.0 Design Features Chapter 5.0 Administrative Controls 4
Part 4, GTS Volume 1, Specifications Part 4, GTS Volume 2, Bases Chapter B 2.0 SLs Chapter B 3.0 LCOs and SRs Part 2, FSAR Chapter 16, TS Section 16.1 Technical Specifications TR-101310-NP, Revision 0, US460 Standard Design Approval Technical Specifications Development
Non-Proprietary NuScale SDAA Part 2 Chapter 16 and Part 4 Review DCA Module is shutdown (keff < 0.99)
All indicated reactor coolant temperatures < 420 °F (minimum temperature for criticality) 5 Significant Changes from DCA to SDA GTS Chapter 1 Definition of MODE 3 - Safe Shutdown SDA Module is shutdown (keff < 0.99)
All indicated reactor coolant temperatures < 345 °F (minimum temperature for criticality)
OR PASSIVELY COOLED Any indicated reactor coolant temperature may be 345 °F
Non-Proprietary NuScale SDAA Part 2 Chapter 16 and Part 4 Review Industry Technical Specification Task Force (TSTF) traveler 554, Rev. 1, approved on December 18, 2020 (ML20324A083) and incorporated into Revision 5 of NUREG-1431, Standard TS Westinghouse Plants, changed the definition - as shown:
LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall. LEAKAGE past seals, packing, and gaskets is not pressure boundary LEAKAGE.
6 Significant Changes from DCA to SDA GTS Chapter 1 (contd)
Definition of Reactor Coolant System (RCS) pressure boundary LEAKAGE
Non-Proprietary NuScale SDAA Part 2 Chapter 16 and Part 4 Review DCA US600 Critical Heat Flux Ratio Correlation Safety Limit NSP2
[1.17]
NSP4
[1.21]
Extended Hench-Levy [1.06]
7 Significant Changes from DCA to SDA GTS Chapter 2 Reactor Core Safety Limits SDAA US460 Critical Heat Flux Ratio Correlation Safety Limit NSP4
[1.21]
NSPN-1
[1.15]
Pressurizer Pressure 2285 psia Pressurizer Pressure 2420 psia Reactor Coolant System Pressure Safety Limit Peak fuel centerline temperature { 4901 - (1.37E-3 x Burnup, MWD/MTU) } °F.
Non-Proprietary NuScale SDAA Part 2 Chapter 16 and Part 4 Review DCA LCOs Omitted in SDAA 3.3.5 Remote Shutdown Station Indication-only monitors in I&C equipment rooms No Type-A PAM variables 3.7.3 In-Containment Secondary Piping Leakage Leak-before-break (LBB) methods not used High energy pipe break exclusion criteria met (consistent with BTP 3-4) 8 Significant Changes from DCA to SDA GTS Chapter 3 LCO Subsection Omissions and Additions Additional LCOs in SDAA 3.5.4 Emergency Core Cooling System Supplemental Boron (ECCS ESB) 3.3.3 Function 1 ECCS hour post reactor trip actuation timer; SR 3.3.3.3 3.6.3 Containment Closure Mode 3 and Passively Cooled; Mode 4 before unseating of upper module assembly from lower containment vessel flange Maintain reactor coolant inventory to ensure adequate core cooling 3.6.4 Passive Autocatalytic Recombiner Meets 10 CFR 50.44(d)
Non-Proprietary NuScale SDAA Part 2 Chapter 16 and Part 4 Review DCA Revision 5 22.a High Containment Water Level 23.a Low RCS Pressure Three reactor vent valves (RVVs)
Two reactor recirculation valves (RRVs)
Inadvertent Actuation Block (IAB) on each valve delays valve opening on ECCS actuation signal until RPV-CNV pressure difference below unblock setting 9
Significant Changes from DCA to SDA GTS Chapter 3 Instrument Functions that Initiate ECCS SDAA Revision 2 (draft) 23.a Low RPV Riser Level (if above 500 °F) 24.a Low Low RPV Riser Level 25.h Low AC Voltage to EDAS Battery Chargers Two RVVs Two RRVs No IAB on RVVs; EDAS DC power ensures ECCS hold function until ECCS actuation signal or reactor trip occurs RRV opening delayed by IAB
Non-Proprietary NuScale SDAA Part 2 Chapter 16 and Part 4 Review DCA Revision 5 7.b High Pressurizer Pressure 13.b High Narrow Range (NR) RCS THOT 16.b High Main Steam Pressure 25.b Low AC Voltage to ELVS Battery Chargers 10 Significant Changes from DCA to SDA GTS Chapter 3 Instrument Functions that Initiate DHRS SDAA Revision 2 (draft) 7.b High Pressurizer Pressure 11.c Low Pressurizer Level 13.b High Narrow Range (NR) RCS THOT 17.b High Main Steam Pressure 22.c High NR Containment Pressure 25.c Low AC Voltage to EDAS Battery Chargers 26.c High Under-the-Bioshield Temperature Low Pressurizer Level Steam Generator Tube Failure High Under-the-Bioshield Temperature High-energy line breaks under the bioshield High NR Containment Pressure Loss of containment vacuum Feedwater System pipe break Inadvertent RVV opening
Non-Proprietary NuScale SDAA Part 2 Chapter 16 and Part 4 Review 3.1.9 Boron Dilution Control
- Added configuration constraints on the Module Heatup System (MHS) to ensure the MHS is never aligned to the CVCS injection line of more than one NuScale Power Module 3.3.1 MPS Instrumentation - accommodating control rod coupling and uncoupling
- Mode 3 Applicability of selected reactor trip and DWSI instrument functions
- when capable of withdrawal of more than one control rod assembly (CRA) 5.5.4 Steam Generator (SG) Program
- 72 effective full power month inspection interval for all SG tubes
- RCS pressure is on the outside of the SG tubes, so the tubes are susceptible primarily to collapse or buckling rather than burst
- Tube plugging criterion of 40 percent through-wall thickness is bracketed as part of COL Item 16.1-1 11 Significant Changes from DCA to SDA GTS Chapters 3 and 5 SDA GTS Improvements Over DCA GTS
Non-Proprietary NuScale SDAA Part 2 Chapter 16 and Part 4 Review NuScale US460 Standard Design GTS and Bases are acceptable because they comply with
- 10 CFR 50.34, Contents of Applications; Technical Information;
- 10 CFR 50.36, Technical Specifications; and
- 10 CFR 50.36a, Technical Specifications on Effluents from Nuclear Power Reactors.
12 Conclusion
Non-Proprietary Presentation to the ACRS Subcommittee Staff Review of NuScales Loss-of-Coolant Accident (LOCA)
Evaluation Model Topical Report (TR 0516-49422-P)
January 15th, 2025 (Open Session) 13
Non-Proprietary Technical Reviewers
- Dr. Shanlai Lu, Lead, NRR
- Dr. Sean Piela, NRR
- Dr. Dong Zheng, NRR
- Mr. Carl Thurston, NRR
- Mr. Ryan Nolan, NRR
- Dr. Syed Haider, NRR
- Dr. Joshua Kaizer, NRR Project Managers
- Getachew Tesfaye, Lead PM, NRR 14 Contributors NuScale LOCA Topical Report Review
Non-Proprietary NuScale LOCA Topical Report Review NuScale submitted Loss-of-Coolant Accident (LOCA) Evaluation Model Topical Report (TR 0516-49422-P), rev. 3, on January 5, 2023. The topical report was formally accepted for review on July 31, 2023 NRC conducted an audit of the topical report from March 2023 to August 31, 2024 57 audit issues were generated, resulting in supplemental information being submitted by NuScale For items not resolved during the audit, 4 RAIs were generated All audit items and RAIs are resolved closed Staff completed review of the LOCA Topical report and issued an advanced safety evaluation to support today's ACRS Subcommittee meeting 15 Overview
Non-Proprietary NRC staff completed the review of LOSS-OF-COOLANT ACCIDENT EVALUATION MODEL topical report. Subject to the closure of the open and confirmatory items noted in the draft SER, and, with eleven limitations and conditions identified, the NRC staff finds the following:
The proposed LOCA analysis methodology is acceptable for meeting the requirements of 10 CFR 50.46 and the associated portions of Appendix K* evaluated in this TR, for evaluation of the ECCS performance in the NuScale NPM-20 for design-basis LOCAs. The proposed LOCA EM is conservative to determine CHF and collapsed liquid level above the reactor core.
The proposed containment response analysis methodology is conservative and acceptable.
The proposed NRELAP5 computer code and the NPM-20 model are acceptable to evaluate the MCHFR for IORV and LOCA events.
- Note that certain portions of Appendix K require exemptions as specified by limitation/condition (post-CHF phenomena).
16 Conclusion NuScale LOCA Topical Report Review
Meeting Title Open Session NuScale Subcommittee on Staff's Evaluation of NuScale Standard Design App Chapters 3, 16 and LOCA Evaluation Topical Repo Attendees Michael Snodderly ACRS (DFO)
Thomas Dashiell ACRS Sandra Walker ACRS Larry Burkhart ACRS R Snuggerud NuScale Jim Osborn NuScale Rob Meyer NuScale Scott Barnes Tyler Beck NuScale Kyle Hoover NuScale Freeda Ahmed NuScale Cindy Williams NuScale Andrea Torres ACRS Dennis Bley ACRS Karl Gross NuScale Dave Petti ACRS Doug B NuScale Shandeth Walton ACRS Kevin Spencer NuScale JJ Utberg NuScale Kevin Lynn NuScale John Fields Wendy Reid NuScale Meghan McCloskey NuScale James Cordes Court Reporter Eric Lantz NuScale Yi-Lun Chu NRR Craig Harbuck NRR Leonard Ward NRR Ron Ballinger ACRS Erin Blumsack NuScale Tom Case NuScale Emily Larsen NuScale Alina Schiller NRR Thomas Griffith - NuScale NuScale Mahmoud -MJ-Jardaneh NRR Matt Sunseri ACRS Andrea Mota NuScale Elisa Fairbanks NuScale Gary Becker NuScale Vicki Bier ACRS Steven Bloom NRR Ricky Vivanco NRR
Walt Kirchner ACRS Robert Martin ACRS Kamal Manoly NRR Allyson Callaway Gregory Halnon ACRS Vesna B Dimitrijevic ACRS Angelo Stubbs NRR Raul Hernandez NRR Warren Erling NRR Gordon Curran NRR Prosanta Chowdhury NRR Thomas Hayden NRR Joy Jiang Milton Valentin NRR Jorge Cintron-Rivera NRR Upendra S. Rohatgi Kenneth Armstrong Stacy Joseph NRR David Benson NuScale Sarah Bristol NuScale Kaihwa Hsu Rob Krsek ACRS Tammy Skov Leslie Terry Jeff Luitjens NuScale Timothy Polich John Bozga Derek Widmayer ACRS Brian Wolf NuScale Marie Pohida NRR Kris Cummings NuScale Sarah Turmero NuScale Ben Bristol NuScale Ata Istar NRR Clint Ashley NRR John Honcharik NRR Greg Makar NRR C Basavarajh NRR Sunwoo Park NRR Shanlai lu NRR Sean Pieta NRR Mike Swim NRR Rebecca Patton NRR Carl Thurston NRR Antonio Barrett NRR Joshua Kaizer NRR Syed Haider NRR Justin Coury RES Cole Takasugi RES
Alfred Krall ERI Getachew Tesfaye NRR
Chapter 16 and Technical Specifications Noteworthy Changes from DCA to SDA Discussion Technical Report TR-101310, US460 Standard Design Approval Technical Specifications Development, Rev 0 describes differences between US600 and US460 Technical Specifications at the time of SDAA submittal.
The reasons for changes are described in general terms, and includes removals, relocations, and new requirements.
LCO 3.1.2 Core reactivity balance surveillance frequency was clarified.
The response to Audit Item A-16.3.1.2-1 revised SR 3.1.2.1 by removing the note associated with adjustment of predicted reactivity values to correspond to measured core reactivity prior to exceeding a fuel burnup of 60 EFPD.
NuScale has no basis for the inclusion of this note other than consistency with the Standard Technical Specifications. The note implied that adjustment of predicted reactivity values is prohibited beyond 60 EFPD. There is no restriction on the timing of the revision of predicted reactivity values.
The revision to SR 3.1.2.1 also removed a note in the frequency column. The note described when the surveillance is to be performed, and was unnecessary.
The Surveillance Frequency Control Program (SFCP) establishes the surveillance frequency.
TS 3.1.9 modified to incorporate additional controls on possible dilution flow paths associated with the Module Heatup System (MHS).
Responses to Audit Items A-16.3.1.9-2 and A-16.3.1.9-3 revised TS 3.1.9 to include:
- New LCO related to MHS flow paths
- Revision to Mode 3 Applicability to include with any dilution source flow path not isolated
- Changes to Actions to address new MHS LCO
- Changes to SR 3.1.9.5 to clarify verification that MHS flow paths to and from cross-connected systems are isolated.
- Supporting Bases changes
The MHS heats the RCS to assist in developing natural circulation through the core before nuclear heat addition.
The MHS is shared among NPMs and, when in service for a module, could represent an inadvertent dilution source for other modules. The revisions to LCO 3.1.9 ensure the modules not being heated by MHS are isolated from the MHS by two closed valves.
TS 3.5.4 modified to address the form of the emergency core cooling system supplemental boron (ESB) pellets and the associated requirements to be specified in the core operating limits report.
Response to Audit Item A-16.3.5.4-1 revised TS 3.5.4 and associated Bases to address the form of boron pellets and the associated requirements to be specified in the core operating limits report.
The pellet dissolution rate depends on the geometric form (dimensions and shape) of the boron pellets.
TS 5.5.4, Steam Generator (SG) Program, revised to update the tube integrity discussion, plugging criterion and inspection requirements.
To determine an appropriate steam generator tube plugging criterion for the US460 design, NuScale performed a finite element analysis specific to the US460 design. TS 5.5.4 was updated to reflect the analysis, and bracket the tube plugging criterion Revisions to inspection requirements increased inspection frequency and specificity.
1 Supplement - ACRS - Loss-of-Coolant Accident Evaluation Methodology Topical Report Review Presentation - Closed Session, January 15, 2025 The staff is providing this supplement to highlight differences between the draft Advance Safety Evaluation Report (ASER) for NuScale, LLC. Topical Report "Loss-of-Coolant Accident Evaluation Model," TR-0516-49422, Revision 3 that was submitted to the Advisory Committee for Reactor Safeguards (ACRS) for review on December 14, 2024 and the version that was published on January 7, 2025 (ML24312A002). These differences do not change any of the staffs conclusions. The staffs presentation during the January 15, 2025 ACRS subcommittees closed session accurately reflected the staffs conclusions in the published version of the ASER.
In summary, the differences include editorial changes, clarifications, and refinement of language in the limitations and conditions. The technical differences are the inclusion of additional supporting evidence of the staffs conclusions, specifically for scaling analysis of the LOCA EM, N-RELAP5 code version use, and containment response analysis methodology (CRAM). The changes are mainly due to additional information submitted by the applicant that supported the staffs conclusions in the draft ASER.
The table below lists the SE sections where changes were made, summary of the changes,
and the slide number where this information was presented during the January 15, 2025, closed subcommittee meeting:
SER Section Summary of Change Presented on Slide #
Section 4.4.2, Phenomenon Identification and Ranking Table Rankings As a result of additional information submitted by NuScale and confirmatory analysis performed by the staff, the staff found the existing PIRT for the in-vessel flow and heat transfer is not impacted by the generation and transport of the small amount of radiolytic gas.
14 Section 4.5.1.5, Helical Coil Steam Generators (HCSG)
The staff confirmed that the DHRS modeling and coupled pool nodalization is sufficient to model the overall decay heat removal responses and heat transfer capability.
15 Section 4.5.1.6, Containment Vessel and Reactor Pool The staff confirmed that the uncertainty in natural convection heat transfer modeling from the CNV and DHRS to the pool due to thermal stratification would not be safety-significant with respect to the containment pressurization and DHRS capacity.
29 Section 4.6, NRELAP Computer Code The staff confirmed code update and basemodel version-to-version benchmark results and determined that the code version update and model changes are acceptable and consistent with this NPM methodology.
9
2 SER Section Summary of Change Presented on Slide #
Section 4.7.5.1, Test Facility The staff confirmed, based on the results of the various assessment sensitivity studies, the applicants conclusion that the NRELAP5 model responses are consistent with physics-based expected results and that there are very negligible effects on the event FOMs.
11-12 Section 4.8.2.7,
((
))
The staff confirmed that heat transfer from the lower head to the reactor pool has a minor impact on the CNV pressure response and that using the ((
)) for modeling heat transfer from the lower hemispherical CNV head does not have any safety-significance with respect to the CNV T/H response.
29 From Section 4.8.3.2.4, Reactor Coolant System Depressurization Scaling The staff confirmed ((
)) and conservatism.
Therefore, the conclusion in the distortion analysis is acceptable.
12 Section 4.8.3.3, Assessment of NuScale Facility Integral Effect Test Data The staff confirmed that the applicants extensive assessments in the LTR with these NIST-2 tests, and the code-to-data agreement is excellent for the figures of merit.
12 Section 4.8.3.4, Evaluation of NuScale Integral Effect Tests Distortions and NRELAP5 Scalability As a result of additional analysis and justification provided by the applicant, the staff confirmed that NIST-2 chronology scaling is maintained as it was in the NIST-1 scaling.
11 Section 7. Limitation and Condition Section 4.5.2, Analysis Setpoints and Trips.
L/C modified to include unless the method is followed that is described in section 5.2 of the LOCA EM TR that models the riser level instrument setpoint based on mixture level in the riser, using one of the approaches described in detail in section 5.2 (not including the application - specific alternate approach). (This addition is also reflected in Section 4.5.2.). Additionally, L/C #4 and #9 were combined into one L/C.
13